IR 05000313/1993032
| ML20059G282 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 01/11/1994 |
| From: | Pellet J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20059G279 | List: |
| References | |
| 50-313-93-32, 50-368-93-32, NUDOCS 9401240085 | |
| Download: ML20059G282 (194) | |
Text
{{#Wiki_filter:._ , APPENDIX U.S. NUCLEAR REGULATORY COMMISSION '
REGION IV
Inspection Report: 50-313/93-32 , 50-368/93-32 C ' Licenses: DPR-51 NPF-6 _; Licensee: Entergy Operations, Inc.
Route 3, Box 137G Ru*.sellville, Arkansas . Facility Name: Arkansas Nuclear One, Units 1 and 2 Inspection At: Russellville, Arkansas Inspection Conducted: December 6-17, 1993 Inspectors: R. E. Lantz, Chief Examiner, Operations Section Division of Reactor Safety ' , S. L. M'Crory, Examiner, Operations Section Division of Reactor Safety Accompanying Personnel: K. Faris, Examiner, Contractor Battelle Pacific NW Labs
N. Maguire-Moffitt, Examiner, Contractor Battelle Pacific NW Labs ' T. Vehec, Examiner, Contractor Battelle Pacific NW Labs Approved: _ lh8 i di Pellet, Chief, Operations Section Date Division of Reactor Safety Inspection Summarv Areas inspected (Unit 2): Routine, announced inspection of the qualifications ! of applicants for operator licenses at the Arkansas Nuclear One, Unit 2, ' facility, which included an eligibility determination and administration of comprehensive written examinations and operating tests.
The examination team ' also observed the performance of on-shift operators and plant conditions incident to the conduct of the applicant evaluations.
The examiners used the guidance provided in NUREG-1021, " Operator Licensing Examiner Standards," ' Revision 7, Sections 201-203, 301-303, and 401-403, issued January 1993, 9401240085 940113 PDR ADDCK 05000313 O PDR, ,
. .. -2-Areas Inspected (Unit 1): No inspection of Unit I was performed.
Results (Unit 2): All nine of the applicants for reactor operator and all eight of the
applicants for senior reactor operator licenses satisfied the requirements of 10 CFR 55.33(a)(2) (Section 1.6).
The reference material provided by the training department for
examination development was. adequate, with noted areas for improvement (Section 1.1).
All applicants passed the written examinations, with scores ranging from
> a low of 82 percent to a high of 95 percent with averages of 87 percent for reactor operator applicants, 89 percent for senior reactor operator ' applicants, and 87.9 percent overall (Section Ic2).
The examiners observed good communication practices from applicants
during the conduct of the examinations and from control room operators during plant walkthr oughs (Section 1.3.1).
In general, applicants did not conduct normal boration control in
accordance with approved procedures during the operating test, resulting l from ineffective training (Section 1.3.1).
! In one instance, component labeling was not as described in approved
procedures and was noted to adversely affect operator performance during the walkthrough examinations (Section 1.3.2), j Results-(Unit 11: Not applicable.
f Summary of Inspection Findinas: There were no findings that were assigned a tracking number identified
during the course of this inspection.
. Attachments: i Attachment 1 - Persons Contacted and Exit Meeting i
Attachment 2 - Simulation Facility Report ,
Attachment 3 - Written Examination Keys !
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-3-DETAILS 1 LICENSED OPERATOR APPLICANT INITIAL QUALIFICATION EVALUATION (NUREG-1021) During the inspection, the examiners evaluated the qualifications of 17 license applicants: 9 for reactor operator (RO), 4 for senior reactor operator (SRO) currently licensed as R0s, and 4 for instant SRO. The inspection assessed the eligibility and administrative and technical competency of the applicants to be issued licenses to opdrate and direct the operation of the reactivity controls of. a commercial nuclear power facility in accordance with 10 CFR Part 55 and NUREG-1021, " Operator License Examiner Standards," Revision 7, Sections 200 (series), 300 (series), and 400 (series).
Further, the inspection included evaluations of facility materials, procedures, and simulatior-m bility used to support development and administration of the exannnations.
These areas were evaluated using the guidance provided in the areas of NUREG-1021 cited above. Additionally, the examination team also observed the performance of on-shift operators and plant conditions incident to the conduct of the applicant evaluations.
After completion of the evaluations, the examiners determined that the nine applicants for R0 licenses and eight applicants for SR0 licenses satisfied the requirements of 10 CFR 55.33(a)(2) and have been issued the appropriate licenses.
Performance results for individual applicants are not included in this report because inspection reports are placed in the NRC Public Document Room as a matter of course.
Individual performance results are not subject to public disclosure.
1.1 Facility Materials Submitted for Examination Development The chief examiner reviewed the licensee's materials provided for development of the examination, which included station administrative and operating procedures, lesson plans, question banks, simulator scenarios, and job performance measures (JPM). The procedures and lesson plans were adequate.
Some JPMs were not current with the latest procedure revision.
The facility bank of written questions, dynamic simulator scenarios, and JPMs was adequate in scope, depth, and variety.
It was used extensively in ~ developing the examinations.
The JPM bank lacked alternate success path JPMs, which required the NRC examiners to develop a larger percentage of the walkthrough section of the operating tests. Additionally, the simulator cause and effect manual was not updated to reflect the in-plant and corresponding simulator annunciator upgrade which was recently installed.
There is no regulatory requirement for a facility to develop and maintain a bank of valid test items (questions, JPMs, and scenarios) for NRC use to develop examinations.
However, because of the significant savings in
-4-development time, the NRC has expressed willingness to use such material if it is available and meets the standards of NUREG-1021.
1.2 Fritten Examinations The examination team developed comprehensive written R0 and SR0 examinations in accordance with the guidelines of NUREG-1021, Revision 7, Section 401.
The examinations consisted of 100 multiple choice questions. During the week of November 22, 1993, members of the facility operations and training departments, under the provisions of NUREG-1021, which require execution of a nondisclosure security agreement, reviewed the examinations at the Arkansas Nuclear One Training Center.
The NRC considers the preadministration review of the examination by the facility as part of the examination development process.
Therefore, the specific comments resulting from that review are not reported or otherwise retained.
The chief examiner incorporated the facility review comments and administered the examinations to the license applicants.on December 13, 1993.
The chief examiner provided the facility training staff with a copy of the "as administered" written examination and answer key along with the preadministration review comments on December 13, 1993, immediately following the completion of the examination by the applicants. The facility reviewed the as-administered examination and provided no additional comments.
One question on the SR0 examination was clarified during the conduct of the examination, and the key was changed to correct a typographical error.
The chief examiner made the appropriate revisions to the examination keys.
Overall, applicants performed well on the written examinations.
Scores ranged from a low of 82 percent to a high of 95 percent with averages of 87 percent for reactor operator applicants, 89 percent for senior reactor operator applicants, and 87.9 percent overall. All applicants passed the written examination.
The chief examiner reviewed applicant performance on individual questions and observe * that the following questions were missed by 50 percent or more of the applicants responding to the question.
The questions are referenced here only by examination 1cvel and question number.
Refer to Attachment 3 for the complete question and answer, , Common questions (SR0/R0): 13/20, 17/24, 23/31, 54/65, 59/68, 89/89 Questions on the R0 examination: 92 Questions on the SR0 examination:
The chief examiner concluded that no specific area of significant knowledge-weakness was apparent in the responses to the above questions.
Therefore, the information is provided to the facility training staff for consideration as feedback into future training need.3 Ooeratina Tests The examiners developed comprehensive operating tests in accordance with the: guidelines of NUREG-1021, Revision 7, Section 301.
The operating tests consisted of two partr, a dynamic simulator scenario portion and a control rooWplant walkthrough portion. The chief examiner previewed and validated the various portions of the operating tests in the Region IV office during the week of November 15, 1993, and onsite on November 22-24, 1993, with the assistance of facility training personnel under security agreement. The examination team administered the operating tests during _the weeks of December 6 and 13, 1993.
1.3.1 Dynamic Simulator Scenarios The examination team evaluated six crews (one consisting of three SR0 instant applicants rotating through three positions, one consisting of one SR0 instant and one R0 applicant, and four consisting of one SR0 upgrade and two R0 applicants) on two or three scenarios (depending on crew composition) using the Arkansas Nuclear One, Unit 2, plant-specific simulation facility. The examiners compared applicants' actual performance during the scenarios with expected performance in accordance with the requirements of NUREG-1021, Revision 7, Section 303, to evaluate applicants' competency on this' portion of the operating tests.
The examination team noted good communication practices among crew members, with only isolated instances of open-ended and informal communications. The crews' use of " peer checks" (a practice where a second operator verifies as correct the component that is about to be operated through a verbal request of ' peer check', and confirmation with ' peer check SAT',) was noted as an excellent practice.
Crew briefings by the SR0 were generally effective and timely, although in one instance the SR0 thoroughly briefed the crew and then conducted an improper shutdown by initiating a reactor trip with emergency core cooling system components disabled (in pull-to-lock), contrary to Procedure 2203.018, " Inadvertent SIAS."
Boration control during power operations was observed to be inconsistent among the applicants and not in accordance with Procedure 2104.003, " Chemical Addition." The applicants developed a practice of maintaining a large value entered into the boric acid makeup (BAM) flow batch controller (2fQIS-4926),, then periodically borating at.a small rate during power operations.
While performing this boration, the operators did not routinely perform Step 2.6 of Supplement 6, "RCS Boration During Power Operation," due to the expected preset value entered on Flow Batch Controller 2FQIS-4926.
This practice was observed to result in ineffective training during the dynamic simulator examinations as follows. One applicant was unable to conduct a normal boration from the BAM tanks to the charging pump suction when an initial value of zero (0) was entered in Flow Batch Controller 2FQIS-4926. The applicant did not perform Step 2.6 of Supplement 6 and reported the controller inoperable, when no malfunctions were present.
A second applicant responded
6- - , similarly, but eventually noticed the zero value and successfully commenced the boration. A third applicant referenced the incorrect procedure while commencing a boration at power, using Supplement 9, "Borating The RCS To Desired Boron Concentration." This third operator also did not perform all steps of Supplement 9 where Flow Batch Controller 2FQIS-4926 was referenced.
All 17 applicants passed this portion of the operating tests.
1.3.2 Walkthrouah Examinations The examination team evaluated each of the R0 and instant SR0 applicants using 10 JPMs relating to tasks within the scope of potential duties of a licensed R0 or SR0 (which included nonlicensed operator tasks outside the control room).
The examination team evaluated the remaining upgrade SR0 applicants on five RO or SR0 tasks each.
The applicants performed some of the tasks in the simulation facility in the dynamic mode.
They simulated (through discussions) the remainder of the tasks in the plant integrated control room and at local operating stations throughout the plant.
Immediately following the performance of each task, the examiners. asked pre-scripted questions relating to the system involved in the task.
The questions solicited "short-answer" responses and permitted the applicants to use operationally controlled references to aid in their responses, unless specifically annotated to require response from memory.
The examiners combined the applicants' task performance and question responses in accordance with the guidelines of NUREG-1021, Revision 7, Section 303, to evaluate performance on this portion of the operating examination.
Overall, the applicants performed adequately. All applicants passed this portion of the operating examination with satisfactory overall performance on systems and tasks.
Each applicant was required to enter controlled access to complete one or more tasks.
Some applicants were unfamiliar with facility escort procedures on entering controlled access with an escorted visitor.
The examiners noted one instance of poor labeling which directly affected the applicants' performance on the respective task. The task utilized Procedure 2105.009, "CEDM Control System Operation," Step 7.3, to place the second motor-generator set in operation, locally at the motor-generator sets.. Several applicants had difficulty completing the task and were generally unfamiliar with local motor-generator set operation.
Cabinet 20163 was not c labeled, and the synchroscope selector and paralleling switches were not labeled as described in the procedure.
The examiners noted that this was an infrequently performed task.
The examiners observed that applicants were very aware of activities in the plant and were quick to report discrepancies or unusual conditions to the control room.
In one instance, an applicant observed and reported a chilled water supply leak to a steam generator blowdown radiation monitor in the Auxiliary Suilding, which was quickly attended to by maintenance.
In a second ,
, . . . . . . i l-7-instance, an applicant observed that a plastic tarp set in the overhead to catch falling debris from overhead. cleaning was loose and quickly reported it to health-physics personnel.
1.4 Observations The examination team observed the performance of on-shift operators and plant conditions incident to the conduct of the applicant evaluations. These observations did not impact the evaluation of individual applicants and are included in this report for information only.
Material condition of the plant was noted as good, with the facility
painting program and shift-to-shift cleanup observed as especially active and effective.
Control room communications were observed during a crew turnover and
diesel generator surveillance prebrief.
In general, communications were formal and effective, consistent with observations made during the dynamic simulator section of the operating tests.
, The examiners questioned the facility trainers on the methods used in
the Unit 2 control room by on-shift operators to conduct a normal boration at power. The trainers confirmed that control room operators perform the boration in the same manner as the examiners had observed of the applicants, as described in Section 1.3.1 above.
This practice does not appear to be in accordance with Procedure 2104.003, " Chemical Addition," Supplement 6, "RCS Boration During Power Operation."
, ~In two instances an examiner was mistakenly restricted from entering a
radiological controlled area that was roped off and posted as requiring only a TLD for entry. The facility escorts (a trainer in one instance, an applicant during his walkthrough in the other) in both instances stated that a self-reading dosimeter was additionally required for entry.
> 1.5 Simulator Fidelity
During the preparation and conduct of the operating tests, the examination team observed no discrepancies in simulator fidelity.
i , ' 1.6 Conclusions The examination team concluded that the performance of all 17 applicants for operator licenses satisfied the requirements of 10 CFR 55.33(a)(2) and recommended that licenses be issued.
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J . =. ,A.. -- .e . -8-In general, the examination team concluded that: Individual applicants and crews performed above average. Communications
and teamwork were noted strengths.
The operations and training practice of maintaining a large value in the
BAM flow controller during power operations was ineffective training, which resulted in boration control methods not in accordance with approved procedures.
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ATTACHMENT 1 1 PERSONS CONTACTED 1.1 Licensee Personnel
- C. Anderson, Manager, Unit 2 Operations
- B. Bement, Manager, Training and Emergency Preparedness M. Blanchard, Simulator Instructor
- R. Edington, Plant Manager (Unit 2)
F. Forrest, Control Room Supervisor J. Hatman, Simulator Instructor
C. Reed, Shift Operations Superintendent D. Russell, Simulator Engineer
- D. Sealock, Supervisor, Simulator Training M. Smith, Shift Engineer C. Simpson, Reactor Operator J. Taylor, Simulator Support.
- J. Wade, Supervisor, Operations Training 1.2 NRC Personnel
- R. Lantz, Reactor Engineer, Region IV K. Faris, Contractor, Battelle Pacific NW Labs N. Maguire-Moffitt, Contractor, Battelle Pacific NW Labs T. Vehec, Contractor, Battelle Pacific NW Labs In addition to the personnel listed above, the examiners contacted other personnel during this inspection period.
- Denotes personnel that attended the exit meeting.
2 EXIT MEETING An exit meeting was conducted on December 17, 1993. During this meeting, the chief examiner reviewed the scope and generic findings of the inspection. The chief examiner did not disclose preliminary results of individual evaluations since they are subject to change during the final review and approval process.
The licensee did not identify as proprietary any information provided to, or reviewed by, the examiner.
The licensee did not state any position on the findings presented during the exit meeting.
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. ATTACHMENT 2 SIMULATION FACILITY REPORT Facility Licensee: Arkansas Nuclear One, Unit 2 Facility Docket: 50-368 Operating Tests Administered at: Arkansas Nuclear One, Unit 2 Operating Tests Administered on: December 7-16, 1993 These observations do not constitute audit or inspection findings and are not, without further verification and review,. indicative of noncompliance with 10 CFR 55.45(b). These observations do not affect 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.
During the dynamic operation of the simulator in support of the operating tests, no previously unidentified simulator fidelity problems were observed.
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,, ATTACHMENT 3 ' WRITTEN EXAMINATION KEYS . I
, . . - - U. S. NUCLEAR' REGULATORY' COMMISSION SITE SPECIFIC EXAMINATION REACTOR OPERATOR LICENSE REGION 4' CANDIDATE'S NAME: FACILITY: - ArkansqA Nuclear One-2' REACTOR TYPE: PWR-CE DATE ADMINISTERED: 93/12/06- ' INSTRUCTIONS TO CANDIDATE: Use~the answer sheets provided to document your answers.
Staple this' cover sheet on top _of the answer sheets.
Points for each-question are' indicated in parentheses after the question.
The passing grade requires a final grade of at_least 80%. Examination papers will be picked up four (4)' hours after the examination starts.
CANDIDATE *S TEST VALUE SCORE % 100.00 % ' TOTALS FINAL GRADE All work done on this examination is my own.
I.have neither_given nor received aid.
Candidate's Signature , e b
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. - ... .~ , . . - - - . _ _.. .. .. . , ! REACTOR OPERATOR Page 2-ANSWER SHEET-Multiple Choice-(Circle or X your choice) l If:you change your answer, write your selection in the blank.
MULTIPLE CHOICE 023 a b c d 001 a b' c d 024 a .b c d-002 a b c d 025 a b c
003 a b c d 026 a b-c d '004 a li c d 027 a.
b c d 005 a b c d 028 a b c d
006 a b c d 029 a b c d , 007 a b c d 030 a b c d 008 a b c d 031 a .b 1: d 009 a b c d 032 La b c d 010 a b c d 033 a b .c d 011 .a.
b c d 034 a b c d
012 a b c d 035 a.
b c d-013 a b c d 036 a b c' d 014 a b c d 037 a b 'c d r 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 j 018 a b.
c d 041 a b c d 019 a .b c d 042 a b c d 020 a b c d 043 a b c d 021 a b c d 044 a b c d , 022 a b c d 045 a b c d , , ' . e = - . . .
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REACTOR' OPERATOR .Page
[ ANSWER SHEET ~ Multiple Choice (Circle or X your choice) If~you change your answer, write your selection in the blank.
1046 a b c-d 069 a b c d 047 a b c d 070 a b c d , -048 a b c d 071 a b c d .j 049 a b c d 072 a b c d
050-a b c d 073 a b c d , 051 a b c d 074 a b c d 052' a b c d 075 a b c d - , 053 a b c d 076 a b c d ^ i 054 ~a b c d 077 a b c d
055 a b c d 078 a b c d 056 a b c d 079 a b c d 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 b c d 083 a b c d 061 a b c d 084 a b c d 062 a b c d 085 a' b c d 063 a b c d 086 a b-c.
d /;
-064: a b c d 087 a b c d-065 .a .b c d 088 a b c d 066 a b c d 089 a b-c d . 067 a b c d 090 a b c d -) 068 a b.
c d 091 a b c d .. - .. .. __., -- . - _ ,
.. . - -. . . . . . REACTOR OPERATOR-Page' 4: ANSWER S'H.E E T- , -Multiple r aoiye (Circle or X'your choice).
_ If-you change your answe'r, write your selection in the' blank.
'092 a b c d , 093 a b c d , 094 a b .c d 095 a b c d-096 a b c d
097 a b c d 098 a b c d ' -099 a b-c d 100 a b c d t 'k , . (********** END OF EXAMINATION **********) , - -. -. - - - -
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NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination the following rules apply: 1. Cheating on the examination means an automatic denial of your application and could result in more severe penalties.
2. After the examination has been completed, you must sign the' statement on the cover sheet indicating that the work is your own and you have not received or given assistance in completing the examination.
This must Ims y done after you complete the examination.
- , 3.
Restroom trips are to be limited and only one applicant at a time may leave.
You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.
4.
Use black ink or dark pencil ONLY to facilitate legible reproductions.
, 5.
Print your name in the blank provided in the upper right-hand corner of the examination cover sheet and each answer sheet.
6. Mark your answers on the answer sheet provided.
USE ONLY THE PAPER PROVIDED AND DO MOT WRITE ON THE BACK SIDE OF THE PAGE.
7.
Before you turn in your examination, consecutively number each answer sheet, including any additional pages inserted when writing your answers on-the ' examination question page.
8. Use abbreviations only if they are commonly used in facility literature.
Avoid using symbols such as < or > signs to avoid a simple transposition error resulting in an incorrect answer.
Write it out.
i 9. The point value for each question is indicated in parentheses after the , question.
10. Show all calculations, methods, or assumptions used to obtain an answer to any short answer questions.
, 11. Partial credit may be given except on multiple choice questions.
Therefore, , ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.
12. Proportional grading will be applied.
Any additional wrong information ' that is provided may count against you.
For example, if a question is j worth one point and asks for four responses, each of which is worth 0.25.
! points, and you give five responses, each of your responses will be worth ' O.20 points.
If one of your five responses is incorrect, 0.20 will be.
> deducted and your total credit for that question will be 0.80 instead of , ' 1.00 even though you got the four correct answers.
, 13. If the intent of a question is unclear, ask questions of the examiner only.
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14. When turning in your examination, assemble the completed examination with examination questions, examination aids and answer sheets.
In addition, turn in all scrap paper.
15. Ensure all information you wish to have evaluated ~as part of your answer is on your answer sheet.
Scrap paper will be disposed of immediately following the examination.
. 16. To pass the examination, you must achieve a grade of 80% or greater.
17. There is a time limit of four (4) hours for completion of the examination.
, 18. When you are done and have turned in your examination,- leave the examination area (EXAMINER WILL DEFINE THE AREA).
If you are found in this area while the examination is still in progress, your license may be denied or revoked.
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QUESTION: 001 (1.00) The'following plant conditions exist: ' 1.
Unit 2 has tripped from 100% power due to a loss of offsite power.
' 2.
Emergency Diesel Generator 2K4A has failed to start.
3.
RCS pressure is 2230 psia.
i 4.
Two (2) CEAs have failed to fully insert into the core.
, 5.
RWT level is 70%. 6.
The Boric Acid flow transmitter down stream of the Boric Acid Makeup ; pumps has just been discovered to be plugged.
' 7.
Reactor operators are performing Emergency Operating Procedure 2202.001, " Standard Post Trip Actions".
, ' WHICH ONE (1) of the following methods is available for emergency boration given the above conditions? a.
HPSI pump 2P-89A through High Pressure Isjection valves.
b.
Gravity feed from RWT through 2CV-4950-2 to the Charging pump 2P36C.
c.
Normal Emergency Boration flowpath through 2CV-4916-2 to the Charging pump 2P36B.
d.
Gravity feed through 2CV-4921-1 to the Charging pump.2P36A.
, , QUESTION: 002 (1.00) WHICH of the following actions should be performed if the Thot input to the Reactor Regulating System (RRS) fails HIGH? ~ , a.
Commence boration.
b.
Take letdown control to MANUAL.
c.
Commence dilution.
, l d.
Take backup heaters to OFF.
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! ., ; -REACTOR OPERATOR' ' Page-8.. ' , . -QUESTION: 003' (1.00) WHICH ONE (1) of the following inputs to CEDMCS is provided by the Plant - Protection System? , a.
AWP (Automatic Withdrawal Prohibit) b.
CWP (CEA Withdrawal Prohibit) c.
AMI (Automatic Motion Inhibit) d.
LGS (Lower Group Stop)
! QUESTION: 004 (1.00) , , WHICH ONE. (1) of the = following would be lost if the : feeder breaker to 480V Bus-2B-7 tripped OPEN?
a.
' Reactor Pressurizer Proportional Heaters.
b.
Hydrogen Recombiner #2M-554 c.
"A" Service Water Pump 2P-4A i ' d.
Motor Generator (MG) Set' .
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l QUESTION: 005 (1.00) WHICH ONE (1) of the following describes the CEDMCS Withdrawal Prohibit (CWP) ! alarm function? ' a.
Initiated by the CEA Position Indication System in response to a PRETRIP condition on DNBR and prohibits ALL movement of. regulating CEAs.
b.
Initiated by the Plant Protection System (PPS) in response to a PRETRIP condition on DNBR and prohibits ALL OUTWARD group movement of CEAs.
c.
Initiated by the CEA Position Indication' System of the Plant Computer in response to OUT OF SEQUENCE / OVERLAP condition and prohibits ALL OUTWARD movement of CEAs.
d.
Initiated by the Plant Protection System (PPS) in response to OUT OF - , SEQUENCE / OVERLAP condition and prohibits ALL movement of CEAs.
QUESTION: 006 (1.00) WHICH ONE (1) of the following conditions is ind.icated by'the BLUE light associated with Reactor Coolant pump (RCP) controls being LIT? a.
The oil lift pump is running.
b.
Oil pressure is NORMAL, c.
An RCP motor over current condition exists.
d.
RCP speed is zero (0) rpm.
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A 1is , j I REACTOR OPERATOR Page 10 . QUESTION: 007 (1.00) WHICH ONE (1) of the following conditions exists if the RED indicating light for the RCP 2P32D is NOT lit? (Assume the bulb has been verified as good).
a.
The 2P32D RCP trip circuit has lost continuity and the ONLY means of tripping the pump is to open the breaker locally.
b.
The 2P32D RCP trip circuit has lost continuity and the local handswitch must be used to trip the pump.
c.
The 2P32D RCP start circuit has lost continuity and the ONLY means of starting the pump is to close the breaker locally.
d.
The 2P32D RCP start circuit has lost continuity and the. local handswitch must be used to start the pump.
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.. . .. . ..... . ._ - ' REACTOR OPERATOR Page 11 - QUESTION: 008 (1.00) 1The following plant conditions exist: 1.
Unit 2 is in COLD SHUTDOWN.
2.
SDC is in service with a solid Reactor Coolant System.
3.
RCS pressure is 300 psia.
4.
RCS temperature is 190 degrees F. and stable.
5.
Steam Generator "2E24A" pressure is 20 psia.
6.
Reactor Coolant pump "2P32A" was started at 0800 hours and secured at 0900 hours to repair an oil leak.
7.
Reactor Coolant pump "2P32B" has been running for one (1) hour.
8.
At 0920 hours Maintenance personnel report that the_ repairs to - Reactor Coolant pump "2P32A" are complete.
WHICH ONE (1) of the following actions, per 2103.006, " Reactor Coolant' Pump Operations", should be taken (at 0920 hours) following a request from maintenance personnel to start RCP "2P32A"? (2103.006, Attachment A is attached).
a.
The RCP can be started after RCS pressure is increased to MINIMUM required per 2103.006, Attachment A.
b.
The RCP can be started with the present conditions at 0930 hours following a 30 minute rest period for the motor, c.
The RCP can be started immediately after establishing a bubble in the pressurizer.
l d.
The RCP can be started after reducing Steam Generator temperature to less than RCS cold leg temperature.
' QUESTION: 009 (1.00) WHICH ONE (1) of the following conditions will AUTOMATICALLY close Letdown Line Stop valve 2CV-4820-2? a.
Safety Injection Actuation Signal, b.
Regenerative HX outlet temp. is 380 degrees F.
c.
Containment Isolation Signal.
d.
Loss of power from 2B61-L3.
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JREACTOR OPERATOR Page 12 l-QUESTION: 010 (1.00) l WHICH ONE. (1). of the following methods is used to control. Shutdown. Cooling System (SDC). purification flow? a.
Manual control of CVCS Backpressure Regulators.2CV-4810/4811.
, b.
Manual control of CVCS Letdown Flow ' Control valves 2CV-4816/4817.
c.
Automatic control of SDC Flow Control valve 2CV-5091.
d.
' Automatic control of SDC Temperature. Control valve 2CV-5093.
, QUESTION: 011 (1.00) , The:following plant conditions exist'. ' 1.
Unit 2 is in COLD SHUTDOWN, 2.
RCS pressure is 400 psia.
3.
T-hot is 190 degrees F.
.i WHICH ONE (1) of the following Engineered Safety Features Actuation' System ' (ESFAS) trip inputs is AUTOMATICALLY enabled when RCS pressure increases to.
436 psia? a.
Low RWT level, i b.
High/ Low Steam Generator level.
c.
Low Steam Generator pressure, d.
HI LOG Power.
I
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. QUESTION: 012 (1.00) WHICH ONE (1) of the following is actuated by LOW Pressurizer pressure signal? a.
Containment Spray System (CSS) b.
Penetration Room Ventilation System (PRVS) c.
Main Steam Isolation System (MSIS) d.
Containment Cooling System (CCS)
QUESTION: 013 (1.00) WHICH ONE (1) of the following is used to designate Category 1 (Post LOCA Qualified) instrumentation? a.
Red dots b.
Green dots c.
Red nameplates , d.
Green nameplates > , ,
. __ - - . -. _ - . _ . .. - . -- . REACTOR OPERATOR Page 14 QUESTION: 014 (1.00)
WHICH ONE (1) of the following is the reason thirty (30) degrees F. is chosen as the setpoint for the Margin To Saturation (MTS) alarm? a.
To ensure net positive suction pressure (NPSH) is maintained for the reactor coolant pumps (RCPs).
b.
To ensure that RCS voiding is prevented.
c.
To ensure that actual subcooling margin is maintained due to instrument errors, d.
To ensure that actual subcooling margin is maintained when the RCPs are stopped.
QUESTION: 015 (1.00) WHICH ONE (1) of the following systems provides cooling to the Containment Cooling System following a design basis Loss Of Coolant Accident (LOCA)? a.
Chilled Water b.
Component Cooling Water c.
Service Water d.
Auxiliary Cooling Water , s
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. - _ - . _. .. . REACTOR' OPERATOR Page.15-QUESTION: 016 (1.00) WHICH ONE (1) of the following conditions will prevent starting Heater Drain Pump 2P8A at 40% reactor power? a.
A Main Feed Pump running and 2CV0742 open.
b.
A Main Feed Pump running and 2CV0742 closed.
c.
B Main Feed Pump running and 2CV0742 open.
d.
B Main Feed Pump running and 2CV0742 closed.
,.. i QUESTION: 017 (1.00) WHICH ONE (1) of the following conditions will AUTOMATICALLY close Feedwater to Steam Generator Isolation valve 2CV-1024-1? , a.
LOW level in Steam Generator "A" b.
HIGH level in Steam Generator "A" c.
MSIS d.
CIAS . - . .
. . _.. . . - _ _ _ . REA'CTOR OPERATOR Page 16 QUESTION: 018 (1.00) The following condition exists: 1.
Unit 2 is at 5% power.
2.
EFAS has actuated.
3.
EFW pump suction is aligned to the Startup and Blowdown Demineralizer effluent.
4.
Suction pressure for the Er.ergency Feedwater (EFW) pumps is.10 psig and DECREASING.
WHICH ONE (1) of the following automatic actions will occur'if EFW pump suction pressure continues to DECREASE to 5 psig? a.
2CV-0789-1 and 2CV-0795-2 will automatically OPEN to supply water from the Condensate Storage Tank.
b.
2CV-0716-1 and 2CV.;-11-2 will automatically OPEN supply water from the Service Water System.
c.
2EFW-0706 will automatically OPEN to supply water from the Condensate Storage Tank.
d.
2EFW-16, Condensate Storage Tank Suction Bypass valve, will automatically OPEN to supply additional water from the Condensate-Storage Tank.
!
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. -. REACTOR-OPERATOR Page 17- ' QUESTION: 019 (1.00) The following conditions exist: Steam Generator "A" Steam Generator "B" 1.
Level 25% & decreasing 20% & increasing 2.
Steam Pressure 725 psig 825 psig - 3.
Feedwater Pressure 1200 psig 1200 psig
WHICH ONE (1) of the following is the current status of the Emergency Feedwater System (EFS) given the above plant conditions? Assume all' plant , systems respond as designed.
>
I a.
EFAS has not occurred but will AUTOMATICALLY actuate when Steam Generator "A" level decreases to 23%. b.
An EFAS trip condition exists and EFW is being supplied to BOTH steam generators.
c.
An EFAS trip condition exists and EFW is being supplied to ONLY Steam Generator "B".
'r d.
An EFAS trip condition exists and EFW is BLOCKED to BOTH steam generators.
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7.. REACTOR OPERATOR Page 18 QUESTION: 020 (1.00) The following conditions exist: 1.
Unit 2 is in HOT STANDBY.
2.
EFW is in MANUAL control.
3.
EFW control valves 2CV-1025-1 and 2CV-1075-1 are throttled to BOTH Steam Generators.
4.
Simultaneous EFAS A#1 and MSIS #1 actuation signals are received'due to an I & C technician error.
WHICH' ONE -(1) of the following is the affect of this error on EFW control valve 2CV-1025-1 and 2CV-1075-1 operation? a.
Valves remain throttled because they are in MANUAL control.
b.
Valves remain throttled because EFAS A#1 and MSIS #1 have NO affect on these valves.
c.
Valves go FULL open due to the EFAS A#1 signal.
d.
Valves go FULL closed due to the MSIS #1 signal.
QUESTION: 021 (1.00) WHICH ONE (1) of the following conditions.will result in AUTOMATICALLY closing: Regenerative. Waste Discharge valve'2CV-4424 (2T-92 discharge to flume) ? a.
Tank inlet valve open.
b.
Discharge line pH reading of 8.3.
c.
Low instrument air pressure of 50 psig.
d.
Waste Inlet Radiation Monitor 2RE-4447 HIGH alarm tri w.
... . . .. ..., - - .-. . , LREACTOR OPERATOR Page 19 QUESTION: 022 (1.00) ' WHICH~ONE (1) of the following discharges to the Drain Collection Header-(DCH)? a.
Blowdown tank drains.
b.
HPSI relief valves.
l C.
LPSI relief valves.
t d.
Containment sump drains.
. , QUESTION: 023 (1.00) WHICH ONE (1) of the following sets of Waste Gas Tank parameters.is acceptable? (2104.010, " Unit 2 Waste Gas Analyzer Operation", Attachment 1"A"
is attached.)
pressure Hydrogen: Oxygen ' , a.
200.psig 401 12% b.
200 psig.
20% 5% c.
250 psig 60% 3% , d.
250 psig 6% 10% . t i A Y .
,_ . . _. . - _., . - _ _ , _.... ' ' i . . REACTOR OPERATOR Page 20 i ., ! QUESTION: 024 (1.00)
WHICH ONE (1) of the following is the cause for an AREA Radiation Monitor to emit a LOW beeping sound? ' a.
Monitor is placed in test.
b.
Monitor has failed LOW.
, c.
A valid HI radiation alarm.
d.
Battery backup is discharged.
i ' QUESTION: 025 (1.00) WHICH ONE (1) of the following RCP discharge piping locations provide penetrations for normal pressurizer spray? a.
RCPs 2P32A and 2P32C.
b.
RCPs 2P32A and 2P32B, c.
RCPs 2P32C and 2P32D.
d.
RCPs 2P32B and 2P32D.
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.... . ~. . . REACTOR OPERATOR' Page 21 > . QUESTION: 026 (1.00) WHICH ONE (1) of the following inputs is provided by Loop 2 hot leg temperature instrumentation? a.
ESFAS bypass permissive for steam generator high/ low water level
trip signals b.
AWP (Automatic Withdrawal Prohibit) in CEDMCS c.
LTOP isolation valve misalignment annunciation when valves'are OPEN-with temperature less than 240 degrees F.
q d.
Steam Dump and Bypass Control System Modulating Opening Signal.
F QUESTION: 027 (1.00) To WHICH ONE (1) of the following locations are the Safety Injection Tanks (SIT) vented when lowering pressure in Mode 3? ' a.
Containment atmosphere b.
Reactor Drain Tank ' c.
Quench Tank d.
Holdup Tank l , i ) ! i '! , l -- -. ..t
_ _ _ ... _.. _ __ _ . _. _ _ _ ._ . - _ _ _. _. _.. . REACTOR OPERATOR-Page 22 ) QUESTION: 028 (1.00) The following conditions exist: l 1.
Unit 2 has tripped due to a small break loss of coolant accident l (LOCA).
2.
The event is ongoing and the Recirculation Actuation System (RAS)- has initiated.
, ' 3.
.RCS pressure has increased to 1500 psia.
, 4.
BOTH High Head Safety Injection (HPSI) pumps continue to operate at full capacity.
. WHICH ONE (1) of the following provides cooling for the HPSI_ pumps? a.
Flow through the pumps to the RCS is sufficient to cool the pumps l under these conditions.
b.
Mini-Recirc flow to the RWT is sufficient to cool-the pumps under' these conditions.
i; ' t c.
Mini-Mini-Recirc flow to the pump suction is sufficient to cool the i pumps for up to thirty eight (38) minutes, i d.
Mini-Recirc flow to the containment sump is sufficient to cool the i pumps for up to thirty eight (38) minutes.
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QUESTION: 029 (1.00) i ' The following conditions exist: 1.
Unit 2 is in COLD SHUTDOWN.
-2.
SDC is in service.
WHICH ONE (1) of the following is the TOTAL affect of Pressurizer Pressure . Channel-1 (2PI4823-1) failing to 500 psia? ! a.
SDC isolation valve 2CV-5084-1 closes.
' b.
SDC isolation valve 2CV-5084-1 closes 100) SIT isolation valves .! 2CV-5003-1 / 2CV-5023-1 open.
- .l c.
SDC isolation valve 2CV-5086-2 closes.
d.
SDC isolation valves 2CV-5086-2 closes AND'EIT isolation valves j 2CV-5043-2 / 2CV-5063-2 open.
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'l QUESTION: 030 (1.00) l The following plant conditions exist: ] 1.
RCS pressure is 2250 psia.
-
2.
Pressurizer Pressure Control System (PPCS) setpoint is 2250 psia.
3.
RCS boron concentration is 500 ppm boron.
4.
Pressurizer boron concentration is 350 ppm boron.
WHICH ONE (1) of the following is the correct method to equalize RCS boron concentration per 2103.005, Pressurizer Operation"? " a.
ENERGIZE all Pressurizer Backup Heaters and place "A" RCP spray valve in MANUAL and OPEN to maintain RCS pressure within + 10 psia of desired setpoint.
~ b.
ENERGIZE all Pressui.er Proportional Heaters and place "A" RCP spray valve in MANUAL and OPEN to maintain RCS pressure within + 10 ~ psia of desired setpoint.
C.
RAISE Pressurizer Pressure Control System'setpoint to 2300 psia.AND energize Pressurizer Proportional Heaters to maintain Pressurizer Spray flow.
, d.
DECREASE Pressurizer Pressure Control System setpoint to 2220 psia AND energize Pressurizer Backup Heaters to maintain Pressurizer Spray flow.
QUESTION: 031 (1.00) Pressurizer level indication on the Remote Shutdown Panel may differ from that of Level Channel 1 as indicated on Panel 2C04.
This is because-the Remote Shutdown Panel receives its signal from ? a.
Level Channel 2 and is NOT temperature compensated.
b.
Level Channel 2 and is temperature compensated.
c.
SPDS which is temperature compensated.
d.
SPDS which is NOT temperature compensated.
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. .. .. -.. , .. . .. . . .. . .. _... -. -,. - - REACTOR OPERATOR Page 25 - QUESTION: 032 (1.00) WHICH ONE (1) of the following signals is used DIRECTLY to generate the-pressurizer level control setpoint? a.
T-cold b.
T-hot.
, c.
Tref d.
Tavg i ' QUESTION: 033 (1.00) WHICH ONE (1) of the following would result from depressing ONE (1)- Manual Reactor Trip pushbutton on panel 2CO3? > a.
TCBs 2 and 6 open and the reactor trips.
b.
TCBs 2 and 6 open but the reactor does NOT trip.
c.
TCBs 2 and'3 open and the reactor trips.
d.
TCBs 2 and 3 open but the reactor does NOT trip.
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REACTOR. OPERATOR.
Page 26 . QUESTION: 034 - (1.00) The following plant conditions exist: 1.
Unit 2 is at 100% power.
2.
RCS pressure is 2405 psia.
- 3.
The Reactor Protection System (RPS) has NOT actuated.
. . 4.
All Reactor Protection System setpoints have been calibrated to > plant design specifications.
WHICH ONE (1) of the following results-from RCS pressure INCREASING to 2450 psia? Assume no additional failures occur.
a.
The HIGH Pressurizer pressure setpoint of the RPS is reached'and Reactor Trip Breakers will OPEN to DE-ENERGIZE the control element , assembly (CEA) holding coils, b.
The HIGH Pressurizer pressure setpoint of the Diverse Scram System
(DSS) is reached DE-ENERGIZING the trip coil to open a contact' ' interrupting output power from both Motor Generator (M3) sets.
c.
The HIGH Pressurizer pressure setpoint of the RPS is reached and the MG set 480 VAC output breakers will OPEN to DE-ENERGIZE the control element assembly (CEA) holding coils.
d.
The HIGH Pressurizer pressure setpoint of the Diverse Scram System (DSS) is reached ENERGIZING the trip coil to open a contact , interrupting output power'from both Motor Generator (M3) sets, t: QUESTION: 035 (1.00) WHICH ONE (1) of the following CEA position alarms is generated by Reed Switch Position Transmitters? a.
CEA GROUP OUT OF SEQUENCE b.
DROPPED CEA c.
LOWER GROUP STOP d.
CEA POWER DEPENDENT INSERTION i , wy a ..,.n,, a - -. --,r.
e _ . . _ - - - - . . . . .., . ~ ~. - , b F REACTOR OPERATOR-Page.27 J
- QUESTION
- 036
.( 1. 0 0 ) 'WHICH ONE (1) of the following is the reason for a FLASHING indication on the ' Saturation' Margin Monitor (SMM) ? I Indicated value is below the saturation margin alarm setpoint.
a.
b.
Indicated value is below 30 degrees F' saturation margin.
c.
Indicated value is erroneous.
j d.
Indicated value is superheated.
. I ! .i QUESTION: 037 (1.00) i WHICH ONE (1) of the following will result in the loss of Containment Spray-
Pump'2P-35A? l a.
Loss of 6900 VAC Buses 2H1.
q b.
Loss of 6900 VAC Buses 2H2.
-; e c.
Loss of 4160 VAC Buses 2A3.
d.
Loss of 4160 VAC Buses 2A4.
i ! ! t . QUESTION: 038 (1.00) !
WHICH ONE (1) of the following is the MINIMUM condition that will INITIATE j actuation of the Containment Spray System? ' ! a.
CIAS actuation, b.
CIAS and SIAS actuation.
! 'c.
Containment pressure at 24.0 psia and:CIAS actuation.
I d.
' Containment pressure at 24.0 psia and SIAS actuation.
i
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. -. _ - . - _ _., . REACTOR OPERATOR-Page 38 QUESTION: 039 (1.00)
WHICH.ONE (1) of the following AUTOMATIC actions take place-following a trip ~ of the Containment Purge Exhaust fan to prevent overpressurization of the containment? ~
a.
Exhaust dampers go'100% open to relieve pressure.
b.
Supply dampers throttle to control pressure.
i c.
Supply' fan trips after a ten'(10) second time delay.
, ' d.
Purge isolation valve 2CV-8284 closes.
! QUESTION: 040 (1.00) The following conditions exist: 1.
Unit 2 has tripped from 100% power one (1) minute ago.
2.
Steam Generator "A" level is 95%. 3.
Steam Generator "B" level is 40%. . 4.
High Level Override Level Selector switch is selected to the BOTH ' position.
WHICH ONE (1) of the following is the expected status of-the Feedwater Control-System (FWCS)? (Assume NO operator action.)
~ a.
Feedwater is isolated to Steam Generator "A" and at 5% flow demand , to-Steam Generator "B".
b.
Feedwater is isolated to BOTH Steam Generators "A" and "B".
! c.
Feedwater is at 5% flow demand to BOTH. Steam Generators "A" and "B".
l , d.
Feedwater is at 5% flow demand to Steam Generator "A" and at' 100% flow demand to "B" Steam Generator.
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__ REACTOR OPERATOR Page 29 t' , QUESTION: 041. '(1.00) The following conditions exist: 1.
Unit 2 is operating at 100% power.
2.
Steam Generator Level selector switch is selected to BOTH.
WHICH ONE (1) of the following will cause the "B" Steam Generator Flow Demand - Signal to DECREASE? a.
"B" Steam Generator Level transmitter fails HIGH.
b.
"B" Steam Generator Steam Flow transmitter fails HIGH.
c.
"B" Steam Generator Feedwater Flow transmitter fails LOW.
d.
"B" Steam Generator Pressure transmitter fails LOW.
> QUESTION: 042 (1.00) ' The following conditions exist: 1.
Unit 2 is operating at 100% power.
2.
Operators are performing the "MSIV Quarterly Partial Stroke Test".
. WHICH ONE (1) of the following will result from holding the EXERCISE switch in the CLOSE position'for one (1) minute? a.
The valve will stroke closed until the GREEN closed light is illuminated, then re-open automatically, b.
The valve will stroke closed until it is 10% closed, then re-open automatically.
c.
The valve will stroke closed until it is 10% closed,' then remain'in l that position until re-opened manually.
, d.
The valve will continue to stroke closed until the EXERCISE switch is placed in the OPEN position.
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t
- -. -- - - -. .. REACTOR OPERATOR Page 30- -, , i QUESTION: 043 (1.00) WHICH ONE (1) of the following is NOT an AUTOMATIC action which results from a' , HIGH alarm trip of Unit 2 Control Room Ventilation Radiation Monitor 2RE-8750-1? a.
Unit 1 Control Room Isolation dampers close.
L b.
Unit 1 Supply fan VSF-8 stops.
, c.
Control Room Emergency Supply fan 2VSF-9 starts.
'd.
Control Room Recirculation dampers open.
QUESTION: 044 (1.00) The following conditions exist:
1.
Unit 2 is in HOT SHUTDOWN.
! 2.
Operators are placing SDC in service.
3.
SDC Heat Exchanger Cross-tie valves 2SI-5A and 2SI-5B are OPEN.
4.
Containment Spray Header MOVs 2CV-5612-1/2CV-5613-2 are OPEN for testing.
WHICH ONE (1) of the following will occur when LPSI pump 2P60A is started? a.
BOTH 2CV-5612-1 AND 2CV-5613-2 will close AUTOMATICALLY.
b.
2CV-5612-1 will close AUTOMATICALLY but 2CV-5613-2 will remain open.
c.
LPSI Header Isolation valve 2SI-6 will close AUTOMATICALLY but 2SI-SB will remain open.
d.
2SI-5A will close AUTOMATICALLY.
,
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REACTOR OPERATOR Page.31 , . QUESTION: 045 (1.00) iWHICH ONE (1). of.the following is a possible source of inventory for the Quench' Tank? a.
Reactor Drain Tank discharge b.
SIT drain lines c.
Service Water System makeup ' d.
RCP Controlled Bleedoff , , QUESTION: 046 (1.00) The'following conditions exist: 1.
Unit 2 has tripped due to a large steam break inside containment.
2.
CSAS has been actuated for one (1) minute.
3.
Sodium Hydroxide pump 2P136A has NOT started.
, WHICH ONE (1) of the following could be a reason the 2P136A pump:is NOT . running? (Assume all systems are performing as designed.)
a.
The pump start time delay has not timed out.
, b '. Sodium Hydroxide pump 2P136A discharge valve'is closed.
' c.
Sodium Hydroxide tank outlet valve.is only 75% open.
d.
Sodium Hydroxide tank level is 50%. , i
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, _ . .. _. . _ - _ , REACTOR OPERATOR Page~32
QUESTION: 047 (1.00) Hydrogen Analyzers must be placed in service within fifteen (15) minutes following a Loss Of Coolant Accident-(LOCA)? This is based on NUREG 1.97 - which requires a hydrogen sample within .? ~ ! a.
thirty (30) minutes of a LOCA and it takes fifteen (15) minutes for the analyzers to warmup.
b.
thirty (30) minutes of a LOCA and'it takes approximately 11 minutes-to transport the gas'to the analyzers, c.
one (1) hour of a LOCA and it takes forty five (45) minutes-for the analyzers to warmup.
d.
one (1) hour of a'LOCA and it takes thirty (30) minutes for the analyzers'to warmup and analyze a sample and approximately 11: minutes to transports.ae gas to the analyzers.
QUESTION: 048 (1.00) WHICH ONE (1) of the following conditions will PREVENT movement of the Refueling Machine? a.
Spreader is NOT extended.
b.
Fuel assembly is in the full up position.
. c.
Suspended ad is at setpoint, , d.
Fuel Hoist is operating.
, -'e e - - - - , e
-.... -- _. REACTOR OPERATOR Page 33 _ QUESTION: 049 (1.00) WHICH ONE (1) of the following Steam Bypass Control System signals /permissives is INOPERABLE should steam line pressure transmitter 2PT-0301 fail LOW? a.
Quick Opening Block signal b.
Automatic Modulation permissive c.
Automatic Motion Inhibit signal d.
Quick opening permissive QUESTION: 050 (1.00) is ~ WHICH ONE (1) of the following conditions renders the Service Water System INOPERABLE per Technical Specification 3.7.3, Service Water System"? " a.
The' Red ESF bus supplies power to both Service Water pump 2P4B and the Auxiliary Cooling Water System (ACW). b.
Service Water pump discharge filter delta pressure is 11 psid.
c.
The Green ESF bus supplies power to both the Service Water pump 2P4B and the 2P4B sluice gate.
d.
Service Water pump suction pressure is 11 psig.
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i I REACTOR OPERATOR Page 34 QUESTION: 051- (1.00) A major Turbine Building Instrument Air header rupture has occurred on Unit 2.
, WHICH ONE (1) of the following is expected to provide Instrument Air to Unit 2 ,from Unit I? a.
Unit 1 Crosstie valve 2SV-3015 AUTOMATICALLY opens.
b.
Unit 1 Crosstie valve 2CV-3004 AUTOMATICALLY closes.
[ ,' c.
MANUALLY open IA TO UNIT 1 CROSS CONNECT. valve IA-51.
d.
MANUALLY open Unit 1 to Unit 2 Crosstie valves 2CV-3015 and 2CV- , 3004.
, QUESTION: 052 (1.00) The following Reactor Coolant Pump conditions exist: "A" RCP "B" RCP 1.
RCP Vapor Seal Pressure 1600 psia 1250 psia 2.
Pump Shaft Vibration 12 mils 14 mils
3.
Pump Frame Vibratica 2 mils 1 mil 4.
RCP Controlled Bleedoff 1 gpm 4 gpm WHICH ONE (1) of the following actions is required per 2203.025, "RCP Emergencies"? ASSUNE the reactor has been tripped.
, a.
Secure "A" RCP due to HIGH RCP Vapor. Seal Pressure.
b.
Secure'"B" RCP due to HIGH Pump' Shaft Vibration.
c.
Secure "A" RCP due to HIGH Pump Frame Vibration.
d.
Secure "B" RCP due to HIGH Controlled Bleedoff flow.
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. . - REACTOR OPERATOR Page 35- . QUESTION: 053 (1.00) WHICH ONE (1) of the following is the MAXIMUM time allowed for operation'of a Unit 2 Reactor Coolant pump following loss of_ Component Cooling Water while in POWER OPERATION per 2203.025, "RCP Emergencies"? a.
1 minute [ b.
10 minutes c.
20 minutes d.
30 minutes QUESTION: 054 (1.00) , The following conditions exist: 1.
Unit 2 is at 100% power.
2.
Total Component Cooling Water flow is 250 gpm to RCP 2P32A.
3.
RCP 2P32A upper seal cavity pressure is 1125 psia.
4.
RCP 2P32A middle seal cavity pressure is 1125 psia.
5.
Pressurizer pressure is 2250 psia.
WHICH ONE (1) of the following describes the condition of RCP 2P32A? ' a.
Low Component Cooling Water (CCW) flow exists.
' b.
Upper seal failure has occurred.
' c.
Middle seal failure has occurred.
d.
Lower seal failure has occurred.
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.. . .- ... REACTOR OPERATOR Page 36 QUESTION: 055 (1.00) The following conditions exist: 1.
Unit 2 has tripped from 100% power and reactor power is DECREASING.
2.
Emergency Boration was initiated at 0800 hours when CEAs #40 and #41 failed to fully insert.
.' 3.
A Pressurizer level control malfunction has resulted in a Pressurizer level of 95% and INCREASING.
4.
Tavg is 565 degrees F. and DECREASING slowly.
5.
Pressurizer pressure is 2400 psia and INCREASING.
6.
Operators are SLOWLY inserting CEA #40 manually and it will be fully 1 inserted at 0815 hours.
7.
CEA #41 remains fully withdrawn and immovable.
8.
Operators desire to terminate RCS makeup to prevent RCS overpressurization.
WHICH ONE (1) of the following conditions will allow termination of-RCS makeup? a.
CEA #40 is fully inserted.
b.
Thirty (30) minutes of Emergency Boration has been completed.
c.
Pressurizer level reaches 100%. d.
Pressurizer pressure reaches 2450 psia.
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- -. - - - .- - - - - - -.. REACTOR. OPERATOR 'Page 37 QUESTION: 056 (1.00) The following conditions exist: , 1.
2P33A is is in PTL for maintenance.
2.
2P33B is supplying loop I CCW.
3.
2P33C is supplying loop II CCW.
' ' 4.
A rupture occurs in loop II CCW, causing pressure to. decrease to less than 80 psig.
WHICH ONE (1) of the following AUTOMATIC actions will result from the above-conditions? a.
CCW Loop I Crossover valves 2CV-5220 and 2CV-5230 close.
b.
CCW Loop II Crossover valves 2CV-5221 and 2CV-5232 close, CCW Loop II Crossover valves 2CV-5221 and 2Cp-5232 open.
i c.
d.
An IMMEDIATE reactor trip on loss of CCW flow to the Reactor. Coolant-pumps.
QUESTION: 057 (1.00) The following conditions exist: 1.
Unit 2 is at 100% power.
2.
Pressurizer spray valve 2CV-4651 has' failed to close on demand.
3.
Operators are taking actions per 2203.028, "PZR Systems Malfunctions", Step 1.
, ! WHICH ONE (1) of the following is the reason the. control switch for 2CV-4651 is placed to OPEN for one (1) second then to CLOSE.for five (5) seconds?. a.
This action OVERRIDES the Master Controller and allows the valve to be closed MANUALLY.
b.
This action. activates the mechanical hammer' mechanism which> assists in seating.the valve.
i ! c.
This action activates the QUICK CLOSE circuitry for the. valve.' d.
This action opens additional CLOSE solenoids which provide greater i air pressure to seat the valve.
, --.. ... - ... ~. _ _, _ ,, _. . . ...,.,, _ _., ,, ._
- . . - .. _ . -. -. . __ _ _ _ - - -... .. .~. REACTOR OPERATOR Page 38' -QUESTION: 058 (1.00) ' - The following conditions exist: 1.
Unit 2 has tripped from an Excess Steam Demand Accident (ESD).
2.
Operators have gone to 2202.005, " Excess Steam Demand".
. WHICH ONE (1) of the following is the reason for tripping one (1) RCP if RCS Tcold is less than 510 degrees F? a.
To decrease heat removal from tre affected loop.
b.
To prevent core uplift with four (4) pumps running.
c.
To decrease affect from pressurized thermal shock (PTS).
d.
To prevent Steam Generator tube stress.
s ,t QUESTION: 059 (1.00) The following conditions exist: 1.
A steam line break exists upstream of the "A" Steam Generator MSIV on Unit 2.
2.
MSIS has automatically initiated.
WHICH ONE (1) of the following conditions could result if a steaming flowpath from the unaffected steam generator is NOT established immediately following dryout of the affected steam generator? a.
Rapid repressurization of the RCS and subsequent Pressurized Thermal Shock (PTS) conditions, b.
Inability to open "B" S/G MSIV due to pressure difference created when affected steam generator reaches dryout. conditions.
i c.
A rapid INCREASE in T-cold of the UNAFFECTED loop resulting in an interruption of natural circulation.
d.
A INCREASE in.the core exit ~ temperatures resulting in an interruption of natural' circulation.
-. . .- -. . . . _ - _ -- . - .. ...
, , . _ ... _ ._ _. -. -.._ . _ _ _.
. .. . . . > REACTOR-OPERATOR Page 39 , . QUESTION: 060 (1.00) WHICH ONE (1) of the following discriminates between a steam line break inside-containment and a small break LOCA? a.
Containment pressure b.
RCS pressure c.
. Containment temperature ! d.
RCS temperature '^ QUESTION: 061 (1.00) - The following conditions exist: . 1.
Turbine roll is in progress.
2.
1800 rpm is selected.
3.
Turbine speed is 1700 rpm and increasing rapidly.
WHICH ONE (1) of the following actions should be performed?
Select'800 rpm on the speed set-rpm selector.
a.
b.
Select Close Valves on the speed set-rpm selector, i i c.
Break vacuum and verify turbine is tripped.
d.
Trip the reactor and verify the turbine is tripped.
,
i
, i . ..
. .. _... .. _ _ _... _... - ._ .. _ _. _.. _ _ .... _ __ _ REACTOR OPERATOR Page 40 . QUESTION: 062 (1.00) WHICH ONE (1). of the following is an indication that natural circulation exists following a. loss of electrical power? (All temperatures are degrees F.)
RCS Subcooling Thot Tcold CETs a.
550 525 540 b.
560 500 570 c.
540 500 530 d.
540 490 560 . QUESTION: 063 (1.00) P-The following conditions exist: 1.
Unit 2 has tripped from 100% power due to.a Loss of Offsite power.
2.
Diesel generators are running with NO Service Water (SW) cooling.
q - WHICH ONE (1) of the following is the MAXIMUM time period that the diesels can run with NO.SW cooling WITHOUT suffering. damage? ' a.
Three (3) minutes b.
Ten (10) minutes C.
Fifteen (15) minutes d.
Diesels can run indefinitely since they. provide theirLown cooling; flow.
i < u i
REACTOR OPERATOR Page 41
QUESTION: 064 (1.00) The following conditions exist: 1.
The control room operator has received a report of a small-fire in the 317' elevation of the Auxiliary Building., 2.
The individual reports the fire is contained in the corner of the 2T20B room where oily rags have been stored incorrectly.
3.
The same individual reports there is no smoke, the area is well ventilated, and he has an ABC Dry Chemical Extinguisher.
4.
The individual has requested permission to extinguish the fire.
WHICH ONE (1) of the following conditions prevent this authorization? a.
The fire is caused by a hazardous chemical.
b.
The potential for smoke exists and the individual has no respiratory equipment.
c.
The extinguisher is incorrect for this type of fire.
d.
The fire is in a radiological control area (RCA).
, f QUESTION: 065 (1.00) WHICH ONE (1) of the following is an indication that is required by Technical Specifications to be displayed at the Remote Shutdown Panel (2C80) ? a. Reactor Trip Breaker Indication b. Steam Generator Pressure c. RCS Subcooled/Superheat Monitor d. Refueling Water Tank Level , h . ~ , ,
_. _ _ _ _ - - ... _ _ _. _ _.
REACTOR. OPERATOR Page 42 ' QUESTION: 066 (1.00) , WHICH ONE (1) of the following operator actions is required during an evacuation of the Unit 2 control room due to a fire in the cable spreading room per 2203.014,_ " Alternate Shutdown"? l ' a.
Trip two (2) Reactor Coolant pumps (one in each loop) and leave two (2) running, b.
Trip ALL four (4) Reactor Coolant pumps.
l c.
Trip the "B" Loop Reactor Coolant pumps.
, d.
Leave ALL Reactor Coolant pumps running if seal water can be established within five (5) minutes.
l . i QUESTION: 067 (1.00) The following conditions exist: 1.
Unit 2 has tripped due to a loss of Feedwater.
2.
Operators have entered Functional Recovery Guidelines.
WHICH ONE (1) of the following is the criteria for going to HR-4, "Once Through Cooling" to provide core cooling? a.
RCS Thot is increasing uncontrollably.
b.
RCS pressure is increasing uncontrollably, c.
BOTH Steam Generator levels decrease to less than 70 inches.
' , d.
RVLMS LVL 03 indicates voiding.
.
4
2
- - . . . .. . - _.
..
. _ .. - _ . _._. _.. _. - ._ .-. _ . _ -. _.
. _ i REACTOR OPERATOR Page 43 QUESTION: 068 (1.00) The following conditions exist: 1.
Unit 2 is at 30% power and increasing at 3% per hour.
2.
CEAs are in MANUAL SEQUENTIAL.
3.
GROUP 5 CEAs were withdrawn from 50 to 60 inches.
4.
At 60 inches Group 5 CEAs continued stepping out.
, WHICH ONE (1) of the following actions are required to be performed FIRST per 2203.003, "CEA Malfunction"? a.
Initiate' emergency boration per 2203.032, " Emergency Boration".
b.
Attempt to stop CEA movement by placing MODE SELECTOR switch.to the.
OFF position.
c.
TRIP the reactor and %; "to 2202.001, Standard Post Trip Actions".
" d.
De-Energize both MG sets by opening breakers 2B712 and 2B812.
, $ QUESTION: 069 (1.00) -The following plant conditions exist: 1.
Mode 3, Trip from 100% power occurred 5 minutes ago.
~ 2.
CEAs #40 and #41 are fully withdrawn.
3.
Charging pump suction piping has ruptured.
WHICH ONE (1) of the following operator actions should be performed under
these conditions? a.
Use auxiliary spray and emergency borate.
- b.
Emergency borate via HPSI by opening 2CVC-115.
c.
Lower RCS temperature and commence a normal boration, d.
Lower RCS pressure and emergency borate via HPSI.
, y ~.r.. , , .- - - - .e . , .,
. . --. .- .. - - . . REACTOR OPERATOR Page 44 , ' ' QUESTION: 070 (1.00) -The'following conditions exist: 1.
Unit 2 tripped from 100% power.
2.
2202.001, " Standard Post Trip Actions", has been entered.
l 3.
The Main Turbine did NOT trip as expected.
WHICH ONE (1) of the following actions should be immediately performed per 2202.001, " Standard Post Trip Actions", to trip /stop the turbine? a.
Manually close MSIVs.
, b.
Manually-open Generator output breakers.
' c.
Manually trip the generator exciter field breaker, ' d.
Locally trip the Turbine from the front standard.
r , QUESTION: 071 (1.00) Considering Procedure 2202.009, " Functional Recovery Procedure", WHICH ONE (1) of the following Safety Functions has the HIGHEST order.of priority?- a.
RCS Inventory b.
Vital Auxiliaries i c.
Core Heat Removal , d.
Containment Integrity '
, y
', t e ? a v ,- , g,,..
. - -. _ ~. .. - _ -. - .. .~ _-- ~. ~ - - . ... . ' REACTOR OPERATOR Page 45
, , QUESTION: 072 (1.00) ! I The following plant conditions exist: 1.
Unit 2 has tripped from 100% power.
2.
SIAS has actuated.
3.
RCS pressure is 1300 psia and DECREASING slowly.
4.
Pressurizer level is 100t.
5.
RCS Thot (Th) is 557 degrees F.
l 6.
Containment pressure is 16 psia and INCREASING slowly.
' WHICH ONE (1) of the following events has occurred? a.
Main Steam line break inside Containment
' b.
Inadvertent SIAS c.
RCS cold leg break
d.
Pressurizer vapor space break ,
. } QUESTION: 073 (1.00) ! WHICH ONE (1) of the following parameters is used to determine if adequate ! core cooling exists following a small break loss of coolant accident (SBLOCA) ' .per 2202.003, " Loss Of Coolant Accident, Safety Function Status Check"?
a.
RCS Thot and average CET temperatures are less than superheated.
! b.
Loop delta T is less than ten (10) degrees F.
l
c.
CET temperatures less than 1000 degrees F.
d.. RCS Tcold is less than ten (10) degrees F. superheated.
~! ! ! l ... - _ _. .. ,_ . . . ,
. _ _ _ - - _ _ . REACTOR OPERATOR Page 46 ' , QUESTION: 074 (1.00) - l The following conditions exist: i i 1.
A large break LOCA condition exists.
! 2.
RCS pressure is 140 psia, j 3.
CETs indicate 267 degrees F.
l 4.
LPSI pump 2P-60A is injecting into the RCS.
i 5.
HPSI pumps 2P-89A and 2P-89B are injecting into the RCS.
I WHICH ONE (1) of the following indicates the lowest acceptable total Emergency Core Cooling System flow per 2202.010, " Standard Attachments"? (References ', attached.)
' a.
750 gpm.
! b.
1455 gpm c.
2125 gpm d.
2900 gpm -, I l QUESTION: 075 (1.00)
The following conditions exist after a Unit 2 reactor trip from 100% power: .
1.
Pressurizer pressure is 450 psia and DECREASING.
[ 2.
Pressurizer level is Ot.
3.
CETs indicate 370 degrees F.
4.
All ESFAS Actuations required have actuated.
WHICH ONE (1) of the following is the MINIMUM operator action for the above conditions per 2202 001, " Standard Post Trip Actions"? (References are provided.)
' a.
Trip RCPs 2P32A and 2P32B.
b.
Trip RCPs 2P32C and 2P32D.
c.
Trip RCPs 2P32A and 2P32C
d.
Trip ALL RCPs.
' l l ! - ' .i - - -. .. - -.- , , - - , v -
. . . . .. - ~ _.- . . ..
. REACTOR OPERATOR Page 47 , QUESTION: 076 (1.00) I WHICH ONE (1) _of the following is the MINIMUM condition that defines a " Harsh" ' containment? a.
Containment temperature is 200 degrees F.
b.
Containment temperature is 150 degrees F. and pressure is greater .; than 8.5 psig.
c.
Containment humidity is saturated.
d.
Containment humidity is saturated and Containment radiation. levels indicate 1 X 10E3 R/hr.
i , & QUESTION: 077 (1.00) WHICH ONE (1) of the following temperature indications is preferred.for monitoring RCS temperature following a loss of Shutdown Cooling flow? a.
LPSI pump discharge monitor (T5096) >
b.
RVLMS ATS j c.
Hot Leg RTDs d.
CETs
.! . ! ? . !
( i , ! J
! , -, -. _.
.
, . .- . .- - -- . REACTOR OPERATOR Page 48
QUESTION: 078-(1.00) WHICH ONE (1) of the following-is the reason MG set supply breakers 2B712 and 2B812 are opened for a minimum of ten (10)_ seconds before re-closing when responding to an Anticipated Transient Without Scram (.ATWS) ? a.
To assure all CEAs have sufficient time to fall into the core.
b.
To allow sufficient time for the TCBs to open.
' c.
To allow sufficient time for the MG sets to coast down and overcome the flywheel device.
d.
To prevent overcurrent trips when breakers are re-closed.
, . QUESTION: 079 (1.00) . The following plant conditions exist:
1.
Unit 2 is in REFUELING.
2.
Start-up channel "A" is out of service.
l 3.
Start-up channel "B" is OPERABLE.
' 4.
Core alterations.are in progress.
WHICH ONE (1) of the following MINIMUM Technical' Specification Action Statements should be implemented? a.
Suspend all operations involving positive reactivity changes.
b.
Determine the RCS boron concentration at least every twelve (12) ) hours.
c.
Immediately evacuate containment until the audible alarm from , Channel "A" is returned to service.
, -
d.
Perform CHANNEL FUNCTIONAL TEST on Channel "B" within one (1) hour.
i } > -! { _ _. -. . _ .. . - , _ _ _. . . . . - -
. -- ... .- .. - REACTOR OPERATOR Page 49 LQUESTION: 080 (1. 0 0 )- The following conditions exist: 1.
Unit 2 has reduced power to 15% due to a Steam Generator tube leak.
2.
Letdown has been isolated.
3.
One (1) charging pump is running.
4.
Pressurizer level is DECREASING slowly.
WHICH ONE (1) of the following is a valid method to determine the amount of primary to secondary leakage? a.
Use Charging flow and Letdown flow mismatch.
b.
Use SG Tube Leak N-16 monitors.
c.
Use Charging flow minus Controlled Bleedoff flow.
f d.
Use Main Steam Line Radiation monitors.
' , QUESTION: 081 (1.00) The following conditions exist: 1.
Unit 2 has tripped from 100% power due to a Steam Generator tube rupture on the "B" Steam Generator.
2.
Operators have throttled HPSI flow to assist in recovery operations.
3.
RCS pressure is 1800 psig.
WHICH ONE (1) of the following conditions will require operators to restore FULL HPSI flow? a.
Margin to saturation is 40 degrees F.
b.
Pressurizer level is 25%. , c.
RVLMS sensor four (LVL 03) indicates WET.
d.
Steam Generator "A" level is 5% and increasing with EFW flow of 500- _ gp _.
.... . -~ _.
REACTOR OPERATOR Page 50 QUESTION: 082 (1.00) The following conditions exist: 1.
Unit 2 has been manually tripped due to a loss of ALL Feedwater.
' 2.
Operators are initiating step 3 of 2202.006, " Loss.of Feedwater" and > .are tripping ALL RCPs.
WHICH ONE (1) of the following is the basis for tripping all Reactor Coolant Pumps as per this step in the procedure? ' j i a.
To reduce heat input into the reactor coolant system.
l b.
To reduce Reactor Coolant System pressure.
c.
To reduce thermal stress to steam generators.
d.
To reduce tube to shell delta pressure in the steam generators.
QUESTION: 083 (1.00) WHICH ONE (1) of the following is the MAXIMUM time'that-the Vital 125 VDC ' batteries 2D-11 and 2D-12 can be expected to carry the vital loads following a loss of the battery chargers? a.
2 hours , i b.
8 hours c.
16 hours d.
24 hours i i
. .. . ... ... .... .. .. p - l REACTOR OPERATOR Page 51 QUESTION: 084 (1.00) -The following conditions exist: 1.
Unit 2 is at 100% power.
2.
Pressurizer level control is selected to Channel "A".
3.
Proportional Heater Bank handswitch is in AUTOMATIC.
WHICH ONE (1) of the following is the affect of a loss of power supply 2Y1 on the Pressurizer Level Control System? a.
Channel "B" instruments FAIL.
. b.
Proportional and Backup Pressurizer heaters are ENERGIZED.
Proportional and Backup Pressurizer heaters are DE-ENERGIZED.
c.
d.
Pressurizer level in; cation at the Remote Shutdown Panel is LOST.
^ QUESTION: 085 (1.00) WHICH ONE (1) of the following is the MINIMUM level required in the Steam Generators in order to sustain Natural Circulation Cooling (NC) ? a.
125 inches b.
200 inches c.
275 inches d.
350 inches _--_____ _-_______ - _- - _
- REACTOR OPERATOR Page 52
. QUESTION: 086 (1.00)
- WHICH ONE (1) of the following valves will fail CLOSED on a loss of Instrument-Air pressure?
a.
2CV-0634, Hotwell Makeup valve b.
2CV-1051, Steam Generator "B" Upstream Dump valve c.
2CV-0740, FW Loop "B" Main Reg valve-d.
2CV-5091, SDC HX Flow Control valve i .- QUESTION: 087 (1.00) - -! The following conditions exist: i 1.
An operator has been assigned to operate the Refueling Machine from 0800 hours until 1300 hours on Monday, 12-06-1993.
2.
Dose rate for the area is 25 mR/ hour.
3.
The operator's exposure for the previous seven days is as follows:
- Monday, 11-29-1993...........
25 mR
- Tuesday, 11-30-1993..........
30 mR
- Wednesday, 12-01-1993........
50 mR
- Thursday, 12-02-1993..........
0 mR , Friday, 12-03-1993........... 25 mR
WHICH ONE (1) of the following is the expected accumulated WEEKLY dose for the.
operator following his present work assignment? a.
125 mR b.
130 mR c.
230 mR - i d.
255 mR , D
-,,
. . . . . - .. . . REACTOR OPERATOR Page 53 QUESTION: 088 (1.00) WHICH ONE (1) of the following is the MINIMUM frequency for, reading-your:Self , Reading Dosimeter' (SRD) while working in a HIGH RADIATION Area? a.
Every sixty (60) minutes b.
Every thirty (30) minutes c.
Every twenty (20) minutes . d.
Every ten (10) minutes ' QUESTION: 009 (1.00) ' WHICH ONE (1) of the following boundaries is challenged the most by.a' chloride excursion in the RCS?
- '
a.
Fuel cladding b.
RCS' letdown piping c.
RCS cold leg piping ' d.
Containment liner , S $
- . - -. .-
, . . . . -... .. - . -.-. . - _-. . .._ . , REACTOR OPERATOR 'Page 54 QUESTION: 090 (1.00) The following conditions exist: 1.
Unit 2 is at 100% power.
2.
An operator is conducting his routine rounds when he discovers a fire door OPEN with scaffolding blocking closure.
' 3.
NO fire watch is present.
WHICH ONE (1) of the following MINIMUM actions must the operator perform due-to these conditions? , a.
Return to his duty station and fill out a Condition Report.
, b.
Report the condition to the Superintendent, Shift Operations.and remain in the area as a fire watch until receiving further instructions.
c.
Perform hourly firewatch patrols until a continuous fire watch is . posted.
I d.
Notify Maintenance Supervisor to remove the scaffolding and remain in the area as a fire watch until the door is closed.
QUESTION: 091 (1.00)
WHICH ONE (1) of the following is an IMPROPER handling of a HIGH pressure gas cylinder? a.
A thirty five (35) pound oxygen cylinder is moved by rolling it on it's bottom edge.
b.
An acetylene cylinder is transported in an OPEN vehicle.
, c.
Grease is applied to a sticky valve on an oxygen. cylinder.
d.
An acetylene cylinder is used in a radiological area.
, .l
.... - _ . -.. . -
.- .. _ .. .. _ _ .. ,
. REACTOR OPERATOR: Page 55 ' QUESTION: 092 (1.00) i WHICH ONE (1) of the following represents a NON-INTENT Change in a' procedure? ' a.
A change in initial conditions.
b.
A modification to setpoints, s c.
Deleting a hold point.
d.
Correction to step sequence.
, QUESTION: 093 (1.00) WHICH ONE (1) of the following controls is placed on keys to the Reactor Building? a.
Controlled by the Superintendent, Shift Operations.
b.
Controlled by the Supervisor, Health Physics.
c.
Controlled by the Manager, Health Physics.
' d.
Controlled by the Plant Manager.
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__.
_ _ _.. _ _ _. _.. - - . ._ .. _.. - REACTOR OPERATOR Page 56 i j QUESTION: 094 (1.00) The-following conditions exist: I 1.
A crew is preparing to conduct MOV testing.
2.
They find a HOLD CARD tagging the breaker OPEN for-the valve they are to test.
- WHICH ONE (1) of the following actions should be.taken next? a.
Install the MOV TEST CARD over the HOLD CARD and proceed with MOV testing.
' ' b.
Proceed with MOV testing under protection of the HOLD C1UU) and a MOV TEST card is not needed.
c.
Install the MOV TEST CARD but contact the Shift Supervisor for further instructions, d.
Contact the Shift Supervisor for authorization to remove the HOLD CARD prior to installing the MOV TEST CARD.
QUESTION: 095 (1.00) WHICH ONE (1) of the following is the PREFERRED method of verifying the correct position of a THROTTLED valve? . a.
Close and re-open the valve the prescribed number of turns.
, b.
Observe the original operator position the valve.
c.
Open and re-close the valve the prescribed number of turns.
d.
Observe control room position indication light.
. t -,,# - - , __ - n -
- .. - = - . ... - i i REACT R OPERATOR Page 57 l - QUESTION: 096 (1.00)
- WHICH ONE (1) of the following methods is used to identify an INACCESSIBLE valve?
a.
By placing a Hold Card on the valve breaker.
b.
By installing bright orange tywrap on the valve.
c.
By painting the valve bright orange.
d.
By installing a special placard on the_ valve breaker.
QUESTION: 097 (1.00) WHICH ONE (1) of the following is the MINIMUM frequency for reading your Self Reading Dosimeter (SRD) while working in a RADIATION Area? a.
Every sixty (60) minutes b.
Every thirty (30) minutes c.
Every fifteen (15) minutes a d.
Every ten (10) minutes i l l j QUESTION: 098 (1.00) WHICH ONE (1) of the following designates a FLOATING STEP in an Emergency Procedure? . a.
An asterisk precedes the step.
b.
An solid square precedes the step, c.
Step number is in brackets, d.
Step number is BOLD.
I f I V , - + - - y-- (w w -,w-: w-my g vy er--vw-- =&-N*'* M D"fTb' I8
~ .. . . . .. - - - - . . , ' REACTOR OPERATOR Page 58-QUESTION: 099-(1.00) .WHICH ONE (1) of the following is the MINIMUM level of approval:needed-to-remove a' NUISANCE. annunciator from service?' a.
Two (2) Licensed Reactor Operators b.
Two (2) Licensed Senior Reactor Operators c.
Shift Engineer AND. Control Room Supervisor d.
Shift Superintendent . t p i QUESTION: 100 (1.00) ~WHICH ONE (1) of the'following activities does NOT require the use of.anL Japproved operating procedure?
a.
Trouble shooting the erratic operation of a EFW valve, b.
Service connection to Service Air System.
, c.
Blowdown of a Safety Injection Tank (SIT).
i d.
Addition of Lithium Hydroxide to the Spent Fuel Pool, ~ 1-
, Li l ' (********** END OF EXAMINATION **********)- ' l < .
.. . . . .. .. t b
. REACTOR [ OPERATOR Page 59 r 'i ANSWER: 001 (1.00) ' c.
[+1.0) REFERENCE: , .1.
2203.032, " Emergency Boration", page 3.
2.
AA52102-003, " Chemical & Volume Control System",. Obj ective. 8.
3.
KA 004000K205 (2.7/2.9) 004010A207 (3.8/3.9) 004010A403 (3.9/3.7) 004010A403 004010A207 004000K205 ..(KA's) , ANSWER: 002 (1.00) b.
[+1.0) , REFERENCE: , 1.
AA52002-014, "Excore Nuclear Instrumentation", Objective 14.3f, page'29.
2.
KA 015000K106 ( 3.1/ 3. 4 ) 015000K106 ..(KA's) . ' ANSWER: ~003 (1.00) b.
[+1.0) , , REFERENCE: 1.
AA-52002-012, " Control Element Drive Control", Objective 12.6, page-15.
l 2.
KA 001000K105 (4.5/4.4) , -001000K105 ..(KA's) -i , ANSWER: 004 (1.00) ' d.- [+1.0) , i
6 -
. . .- -. - - - - . . .. . . .
REACTOR _ OPERATOR 'Page-601 REFERENCE: 11.
AA-52002-012, " Control Element Drive Control", Objective 12.3, page 8.
, ! 2.
KA 001050K202 ( 3.1/ 3 ~. 5 ) , 001050K202 ..(KA's) , . ANSWER: 005 (1.00) b.
(+1.0) REFERENCE: 1.
AA-52002-012, " Control Element Drive Control", Objective 12.6,'page-16.
.2. KA 001000K105 (4.5/4.4) 001000K105 ..(KA's) , 3: ANSWER: .006 (1.00) d.
[+1.0) t REFERENCE: 1.
AA52002-001, " Reactor Coolant System", Objective 1.6, page 39, 2.
KA 003000A304 (3.6/3.6) , 1003000A304 ..(KA's) , ANSWER: 007 (1.00) a.
[+1.0) i ! REFERENCE: ' 1.
AA52002-001, " Reactor Coolant System", Objective 1.6, page 41.
2.
KA 003000A402 (2.9/2.9) 003000A402 ..(KA's) 'l I - -. ,
. - . -.. . - '! REACTOR-, OPERATOR Page'61 . ' ANSWER: 008 (1.00) a :.. [+1.0) REFERENCE: i ' 1.
2103.006, " Reactor Coolant Pump Operations", Attachment A, page 10.
-2.
AA52002-001, " Reactor Coolant System", Objective 1.7, page 42.
, 3.
KA 003000G005 (3.4/3.8), 003000G010 (3.3/3.6), 003000G013 (3. 6/3. 7)
003000G013 003000G010 003000G005 ..(KA's) ANSWER: 009 (1.00) , a.
[+1.0]
REFERENCE: i-1.
AA42002-003, " Chemical and Volume Control", Objective 3.3, page 10.
~2.
KA 004000G007 (3.3/3.3), 004010K403 ( 3.1/3. 6 ) . 004000G007 004010K403 ..(KA's) < ANSWER: 010 (1.00) a.
[+1.0] REFERENCE: 1.
AA42002-003, " Chemical And Volume Control", Objective 3.3h, page 16.
2.
KA 004010K101- (3.4/3.9) 004010K101 ..(KA's) ' ANSWER: -011 (1.00) a.
[+1.0) . ' .-. _ - . _ _
- , REACTOR-OPERATOR .Page 62.
-REFERENCE: 1.
AA-52002-013, " Engineered Safety' Features Actuation System", Objective 1 13.1F, page 11.
12. KA 013000A101 (4.0/4.2) 013000A101 ..(KA's) , , ANSWER: 012 (1.00) d.
(+1.0) REFERENCE: 1.
AA-52002-013, " Engineered Safety Features Actuation System", Obj ective - 13.2B, page 16.
- 2.
KA 013000A301 (3.7/3.9) , 013000A301 ..(KA's) , , RNSWER: 013 (1.00) b.
(+1.0) . REFERENCE: 1.
AA52002-014, "Excore Nuclear Instrumentation", Objective 14.2D, page 10.
2.
101 015000G004 (3.4/3.4) 015000G004 ..(KA's) ANSWER: 014 (1.00) ' c.
(+1.0) . > > , , , ~ , -
-.. - 1 REACTOR OPERATOR Page-63 , i REFERENCE: ~ e L I' -1.
1AA52002-037, " Saturation Margin Monitor", Objective 37.8, page 9.
12 ~. KA 017000G010 (2.6/2.9) 017000G010 ..(KA's) ' ~ ANSWER: 015 (1.00) c.
[+1 0) ' REFERENCE: 1.
STM-2-09, " Containment Cooling and Purge' System",-'page'1.
2.
AA52002-027, " Service Water System",. Objective 27.2J.
3.
KA 022000K101 (3.5/3.7) t 022000K101 ..(KA's) . - ANSWER: 016 (1.00)' d.
_ [ + 1.' 0 ] REFERENCE: , , 1.
AA32003-021, " Condensate and Feedwater System", Objective 21.6, page 41.
.2.
KA 059000G010 (2.9/2.9) , 059000G010 ..(KA's)
ANSWER: 017 (1.00) c.
[ +1. 0] I .- h t I ,
- ,- .- ,. . ~, - - - -
. - - . . , REACTOR OPERATOR.
Page'64 . REFERENCE: 1.
AA32003-021, " Condensate and Feedwater System", Objective 21.3h, page 31.
- 2.
. }Ui 059 000K419 (3.2/3.4) ' >
059000K419 ..(KA's)' ANSWER: 018 (1.00) b.
[+1.0] REFERENCE: 1.- AA-42002-021, " Emergency Feedwater System", Objective 21.2B, page 6.
2.
KA 061000K401 (3.9/4.2) , 061000K401 ..(KA's) - ANSWER: 019 (1.00) c.. [+1.0] REFERENCE: 1.
AA-52002-013, " Engineered Safety Features Actuation System", Objective ' 13.2D, page 20.
2.
KA 061000K414 (3.5/3.7) 061000K414 ..(KA's) . ANSWER: 020 (1.00) > c.
[+1.0] t ! d
i . - . -.
REACTOR' OPERATOR- 'Page 65 .. ' REFERENCE: 1.
AA42002-021, " Emergency'Feedwater System", Objective 21.3C, page 14.
2.. KA 061000A303 (3.9/3.9) 061000A303 ..(KA's)
ANSWER: 021 (1.00) d.
(+1.0) ' REFERENCE: 1, AA-52002-033, "Radwaste System", Objective 33.4, page 27.
2.
KA 068000A302 (3.6/3.6) 068000A302 ..(KA's) ANSWER: 022 (1.00) a.
[+1.0]
- REFERENCE:
- 1.
STM-2-52, " Liquid Waste System", page 5.
2.
AA-52002-033, "Radwaste System", Objective 33.4.
3.
KA 068000K107 (2.7/2.9) 068000K107 ..(KA's) ANSWER: 023 (1.00) c.
[+1.0) . .l l i '
_ . - . ,. REACTOR OPERATOR .Page 66 , REFERENCE: 1.- 2104.010, " Unit 2 Waste Gas Analyzer Operation", Attachment "A", page 23.
2.
AA-52002-033, "Radwaste System", Objective.33.4, page 32.
2.
KA' 071000K504 (2. 5/3.1),. 071000A429 '(3. 0/3. 6) 071000A429 071000K504 ..(KA's)
ANSWER: 024 (1.00) d.
[+1.01 REFERENCE: 1.
AA-52002-018, " Radiation Monitoring System", Objective 18.5, page-9.
2.
KA 072000G010 (2.8/3.0) , 072000G010 ..(KA's)
- ANSWER:
025 (1.00) b.
[+1.0] ' REFERENCE: 1.
AA52002-001, " Reactor Coolant System", Objective 1.1,-page 8.
2.
KA 002000K109 (4.1/4.1). 002000K109 ..(KA's) ANSWER: 026 (1.00) a.
[+1.0] > , REFERENCE: '1.
AA52002-001, " Reactor Coolant System", Objective 1.2, page 10.
2.
KA 002000K108 (4.5/4.6) 002000K108 ..(KA's) _ _ . ,
. _ . . . _ _ . _ . _.. . _ _ . ~ ' 1 REACTOR-OPERATOR- 'Page 67.
' , 1 ANSWER: - 027 (1.00) a.
(+1.0) REFERENCE: l l 1.
.AA52002-004, " Emergency Core System", Objective 4.9A, page 17.
~- 2. KA 006020A107 (3.5/3.7) 006020A107 ..(KA's) , - ANSWER: 028 (1.00) c.
[+1.0) . > .. REFERENCE: 1.
AA52002-004, " Emergency Core System", Objective 4.4, page 9.
2.
KA 006000K603 (3.6/3.9) - 006000K603 ..(KA's) ANSWER: 029-(1.00) ' a.
[+1.0] REFERENCE: ' 1.
AA52002-001, " Reactor Coolant System", Objective 1.11, .2.
KA 010000K403 (3.8/4.1) ' pages 56-57.
. 010000K403 ..(KA's) . ANSWER: 030 (1.00) -, a.
[+1.0} , 4 t -. r -- , - e n r . e, o
. _.. _ _.
. ~.. _. . _... _...... .~ _ .. -... _.. _... _. _..... ! REACTOR OPERATOR Page168-REFERENCE:
! 1.
2103.005, " Pressurizer Operation", page 9.
' 2 '. AA52102-002, STG2A, Objective 1.1, Attachment A, page 32 of 45, - ' 3.
AA52002-001, " Reactor Coolant System", Objective 1.11,.page 60.
4.
KA 010000A402 (3.6/3.4), 010000A101 (2.8/2.9) 010000A101 010000A402 ..(KA's) ! . ANSWER: 031 (1.00) a.
[+1.0} REFERENCE: 1.
AA52002-001, " Reactor Coolant System", Objective 1.9, page 50.
- I 2.
KA 011000K605 ( 3.1/ 3. 7 ) ' ' 011000K605 ..(KA's) ANSWER: 032 (1.00) d.
[+1.0] REFERENCE: 1.
AA52002-001, " Reactor Coolant System", Objective 1.9, page 52.
2.
KA 011000K404 (3.0/3.3) 011000K404 ..(KA's) ANSWER: 033 (1.00) b.
[+1.0) j .1 REFERENCE: -l l 1.
AA52002-006, " Reactor Protection System", Objective 6.8 page 16 ' , 2.
KA 012000A401 (4.5/4.5) 012000A401 ..(KA's) . . - -. - - - ..
. - , _ .. . -.. _-. t REACTOR. OPERATOR Page 69.
. . .' ANSWER: 034 (1.00) ' 'd.
[+1.0]
. REFERENCE: 1.
AA-52002-043, " Diverse Scram System", Objective 43.10,- page 8.
2.
KA 012000K201 (3. 3/3. 7),. 012000K107 (3.2/3.2) ! 012000K103 (3.7/3.8) 012000A407 (3.9/3.9) . 012000A407 012000K103 012000K107 012000K201 ..(KA's)
ANSWER: 035 (1.00) , b.
[41.0) . ' REFERENCE: 1.
STM-2-02, " Control' Element Driv 6 Control", page 2.
2.
AA-52002-012, " Control Element Drive Control", Objective 12.2.
3.
KA 014000K101 (3.2/3.6) 014000K101 ..(KA's) i ANSWER: 036 (1.00) , c.
[+1.0) , ' REFERENCE: 1.
AA52002-037, " Saturation Margin Monitor", Objective 37.6, page 8.
2.
KA 016000A302 (2.9/2.9) , . 016000A302 ..(KA's)
!
' ANSWER:- 037 (1.00)
c.
[ + 1. 0) ! I I .j-I , _.. _ _... . . __ - '
.-- . e.
_-_ .. _ _ _ .-. . . .; - REACTOR OPERATOR ~ Page 70-REFERENCE: -1.
AA42002-007, " Containment Spray System", Objective 7.4.
^2.
STM 2-08, " Containment Spray System", page 3.
, =3.
KA 026000K201 (3.4/3.6)
, 026000K201 ..(KA's) ANSWER: 038 (1.00) d.
[+1.0) REFERENCE: I 1. - AA-52002-013, " Engineered Safety Features Actuation System", Objective; - 13.2D, page 15.
. I 2.
KA 026000A301 (4.3/4.5), 026000K101 (4.2/4.2) 026000A301 026000K101 ..(KA's) ANSWER: 039 (1.00) . ! c.
[+1.0] REFERENCE: ' 1.
STM-2-09, " Containment Cooling and Purge System", page 7.
2.
KA 029000K104 (3.0/3.1) 029000K104 ..(KA's) i i ' ANSWER: 040 (1.00) ,. . a.
[+1.0]
.
? s pix - %
., , . ,.- . , -. ... . -. -. . . - -. ~.. , ~ . REACTOR' OPERATOR Page 71 REFERENCE:
1.
AA52102-015, "Feedwater Control System", Objet ive 4, page 18.
2.
KA 035010A301 (4.0/3.9) , 035010A301 ..(KA's) ANSWER: 041 (1.00) , . a.
[+1.0) ' REFERENCE: 1.
AA52102-015, "Feedwater Control System", Objective 7, page 27.
2.
KA 035010A203 (3.4/3.6) 035010A203 ..(KA's) ' ANSWER: 042 (1.00) d.
[+1.0)
REFERENCE: 1.
STM-2-15, " Main Steam", page 4.
2.
2106.016,. " Condensate and Feedwater Operation",~ Supplement 1, page 116.
3.
KA 039000K408 (3.3/3.4) 039000K408 ..(KA's) ANSWER: 043 (1.00) d.
[+1.0) . '
s , ( ,- _ . .
- . . . ..... . . s i - REACTOR' OPERATOR-Page=72 -~ , JREFERENCE:
fl. AA52002-032, Control-Room and'ESF. Equipment Ventilation. System", " Objective 32.2,Lpage'5.
' '
- 2.
.KA 073000G007 (2.9/3.0)- 073000G007 ..(KA's) ] D , ANSWER: 044 (1.00) b.
[+1.0) ! -REFERENCE: , 1.
AA52002-044, Shutdown Cooling System", Objective 44.2, page 7.
" 2.
KA 005000K110 (3.2/3.4) .005000K110- .(KA's) . , ANSWER: 045 (1.00)- d.
[+1.0]' , P REFERENCE: ' ~ 1.- AA52002-001, " Reactor Coolant System" Objective 1.8, page 62.
, 2.
KA 007000A301 -(2.7/2.9) 007000A301 ..(KA's)
ANSWER: 046 (1.00) c.
[+1.0)
!
. P .m - -. ,
_.
. . - _ - .. _ -. .;~ _.. _. -. _. _, , REACTOR-OPERATOR Page 73-
I
-REFERENCE: 2?. AA42002-007, " Containment Spray System", Objective 7.6, page 20, 2.
KA 027000G007 (3.2/3.4)
I ' .027000G007 ..(KA's) , ANSWER: 047 (1.00) , b.
[+1.0) ! REFERENCE: L 1.
AA52002-050, " Containment Combustible Gas Control", Objective 50.7,:page .21.
. ,
2.
KA 028000G010 (3.0/3.2) 028000G010 ..(KA's)
' ANSWER: 048 (1.00) s d.
[+1.0) . REFERENCE: 1.
AA-52002-006, " Fuel Handling Equipment",. Objective 26.3, Attachment A, page 1.
. , 2.
KA 034000K402 (2.5/3.3) , . 034000K402 ..(KA's) . ANSWER: 049 (1.00) b.
[+1.0) , l
l .) .
_... _ _ _ _ _ _ _ _. _. .. _. _. _. _. _ - , REACTOR OPERATOR Page 74 ~ REFERENCE: ' 1.
AA52002-011,'" Steam Dump &. Bypass Control System",- Objective'11.2.
2.
SDBCS, pages 7 & 9.
' 3.
KA 041020A102. (3.1/3.2) 041020A102 ..(KA's) . ANSWER: 050 (1.00) b.
[+1.0) i REFERENCE: ' i 1.
AA-52002-027, " Service Water", Objective 27.6,,page 29.
2.
KA 076000G005 (2.8/3.2), 076000G010 (2.7/2.9)
,- i ' 076000G010 076000G005 ..(KA's) ANSWER: 051 (1.00) d.
[+1.0] REFERENCE: 1.
2203.021, " Loss of Instrument Air", page 2.
2.
STM-2-48, " Instrument Air System", page 3.
3.
KA 078000K303 (3.0/3.4) 078000K303 ..(KA's) ANSWER: 052 (1.00) a.
{+1.0) REFERENCE: 1.
2203.025, "RCP Emergencies" Attachment D, page 14.
2.
KA 000015G010 (3.4/3.4) 000015G010 ..(KA's) _
G , REACTOR OPERATOR Page 75
, -ANSWER: 053' (1.00) , -b.
[+1.0] REFERENCE: 1.
2203.025, "RCP Emergencies", Attachment D, page 14.
2.
KA 000015A210 (3.7/3.7) 000015A210 ..(KA's) '
. ANSWER: 054 (1.00) t c.
[+1.0] - - I REFERENCE: 1.
2203.025, "RCP Emergencies", page 5.
' 2.
KA 000015A122 (4.0/4.2) , 000015A122 ..(KA's)
' ANSWER: 055 (1.00) a.
[+1.0) REFERENCE: j 1.
AA52002-014, " Reactor Trip Recovery", Objective 14.5, page 6 i 2.
KA 000024K302 (4.2/4.4) l 000024K302 ..(KA's)
'l i ANSWER: 056 (1.00) , C.
[+1.0) < l l .;
REACTOR OPERATOR Page 76-REFERENCE: L '1.
2104.028,-" Component Cooling Water System Operation", page 6.
2.
AA52002-030, " Cooling Water System", Obj ective.30.2H,.pages 15-17.
-3.
KA 000026A201 (2.9/3.5) 000026K301 (3.2/3.5), 000026A102 (3.2/3.3) 000026A102 000026K301 000026A201 ..(KA's)- ANSWER: 057 (1.00) b.
[+1.0] REFERENCE: .1.
AA42102-002, " Pressurizer Pressure and Level Control System", Q10.,.page 6.
~2.
KA 000027K303 (3.7/4.1) . 000027K303 ..(KA's) ANSWER: 058 (1.00) b.
[+1.0] REFERENCE: 1.
AA52003-009, " Excess Steam Demand", Objective 09.2, page 15 2.
KA 000040K304 (4.5/4.7) 000040K304 ..(KA's) . ANSWER: 059 (1.00) a.
[+1.0) - ' i
- REACTOR OPERATOR
'Page 77 ~ REFERENCE: 2.. - AA52003-009, " Excess. Steam Demand", page.6.
2.
KA 000040K106 (3.7/3.8).
000040K106 ..(KA's) -ANSWER: 060 (1.00) d.
[+1.0] REFERENCE: 1.
2202.010, " Standard Attachments", Attachment 20, page 60.
2.
KA 000040A203 (4.6/4.7) 000040A203 ..(KA's)
- (NSWER:
061 (1.00) b.
[+1.0) E REFERENCE: 1, 2203.019, " Loss Of Condenser Vacuum", page 4.
2.
KA 000051A202 (3.9/4.1) -000051A202 ..(KA's) . ANSWER: 062 (1.00) c.
{+1.0) ' REFERENCE: 1.
AA52003-011, " Loss of Offsite Power", Objective 11.6, pages 20-21.
- 2.
KA 000055A202 (4.4/4.6)' 000055A202 ..(KA's) . U . ... -. , -. -. ,. _ _ . . . . .. _
w - a.
n<w.
L ! REACTOR' OPERATOR Page 78 l i IANSWER: 063 (1.00) a.
[+1.0) , . REFERENCE: - .1.
'AA52003-012, " Station Blackout", Objective 12.5', page 14.
2.
KA 000055G007' (3. 6/3. 7) 000055G007 ..(KA's) . ANSWER: 064 (1.00) d.
-[ +1. 0) ' REFERENCE: ' 1.
2203.034, " Fire Or Explosion", page 2.
i , 2.
}UL 000067K102 (3.1/3.9)
000067K102 ..(KA's)
i .I ANSWER: 065 (1.00)
, lb.
[+1.0) -j , REFERENCE: ! '1.
T/S.3.3.3.5, Table 3.3-9, p. 3/4 3-37.
' ' 2.
.L4 000068K201 (3.9/4.0) I 000068K201 ..(KA's) ' ' . j ' ANSWER: 066- (1.00) J b.
[+1.0) ~
.... ,.. _ ._,
- -_ . i REACTOR OPERATOR Page 79-i
. REFERENCE: 1.
2203.014, " Alternate Shutdown", page 3.
2.
KA 000068G010 ( 4.1/ 4. 2 ) . 000068G010 ..(KA's) , i .RNSWER: 067 (1.00)
c.
[+1.0]
REFERENCE: 1.
2202.009, " Heat Removal Decision Tree", page 1.
I 2.
KA 000074G011 (4.5/4.6) .i ' 000074G011 ..(KA's) i ! ANSWER: 068 (1.00) b.
(+1.0] REFERENCE:
2203.003,."CEA Malfunction", page 2.
.i 2.
AA-52002-012, " Control Element Drive Control", Terminal Objective, page ~
2.
.3.
KA 000001G010 (3.9/4.0) I ! 000001G010 ..(KA's) I i . ANSWER: 069 (1.00): e d.- [+1.0]
a i . i !
3 . _ __ -- ....
-. .. -...,..- . .... .... F . REACTOR. OPERATOR .Page 80 ' REFERENCE: j .- 1.
2203.003, "CEA Malfunction", page 2.
j 2.
AA-52002-012, " Control Element Drive Control", Terminal Objective, page 2.
3.
KA 000003G011 (4.0/4.1) .j 000003G011 ..(KA's) , i ANSWER: 070 (1.00) a.
[+1.0) - REFERENCE: , 1.
.2202.001, " Standard Post Trip Actions", page.4.
2.
KA 000007K301 (4.0/4.6) i 000007K301 ..(KA's) . ' ANSWER: 071 (1.00) b.
[+1.0] REFERENCE: 1.
2202.009, " Functional Recovery Procedure", page 11.
- 2.
KA 000007G012 (3.8/3.9) 000007G012 ..(KA's)
ANSWER: 072 (1.00) d.
[+1.0) 'T REFERENCE: ' 1.
AA52002-015, " Loss of Coolant Accident EOP", Objective 15.4, page 17.
2.
KA 000008G011 ( 4. 0 / 4.1 ) + 000008G011 ..(KA's) i . > t f
..... ... m._ .._. . _ _ _ _ _ _ . _ _ ._ _ _ _ _ _ _ REACTOR OPERATOR Page-81 j ANSWER: 073 (1.00) . a.
[+1.0] -l REFERENCE:
1.
2202.003, " Loss Of Coolant Accident, Safety Function Status Check", page.
! 72.
2.
KA 000009A239 (4.3/4.7).
000009A239 ..(KA's) ! ! -ANSWER: 074 (1.00) , c.
[+1.0) . i REFERENCE: 1.
2202.019, " Standard Attachments", Exhibits 2 & 3, pages 90 & 91.
2.
AA52002-015, " Loss Of Coolant Accident EOP", Objective 15.15, page 42, 3.
KA 000011A210 (4.5/4.7) , 000011A210 ..(KA's) i
!
' ANSWER: 075 '(1.00) ' c.
[+1.0) REFERENCR-1.
2202,1'01, " Standard Post Trip Actions", page 8.
' -2.
RFe 2. 010, " Standard Attachments", Attachment 1 and 2,.pages 3 and 4.
3.
Zh 000011G010 _(4.5/4.5), 000011A103 (4.0/4.0) . ., , 000011A103 000011G010 (KA's) i r ANSWER: 076 (1. 0 0 )-
, a.
[+1.0) ,
l > . ' - ' ~ - "- . . .
REACTOR: OPERATOR Page,82 ' REFERENCE: ~1.
.2202.003, " Loss Olf Coolant Accidents, page 2.
a 2.
KA 000011G007 (3.7/3.9) 000011G007 ..(KA's) ANSWER: 077 (1.00) d.
[+1.0] REFERENCE: 1.
2203.029, " Loss of Shutdown Cooling", page 3.
2.
KA 000025A112 (3.6/3.5) ~ ,.. i 000025A112 ..(KA's) ANSWER: 078 (1.00) -c.
[+1.0) ' REFERENCE: -1.
STM-2-02, " Control' Element Drive Control",-page 4.
2.
KA 000029K312 (4.4/4.7) 000029K312 ..(KA's) ' . ANSWER: 079 '(1. 00 ). a.
[+1.0) , , r REFERENCE: -;
1.
Unit 1 TechnicalcSpecifications_3.9.2, " Refueling Operations",-page-3/4 '! 2.
KA b00032G008 - (2.'8/3 -. 3 ) . 000032G008 ..(KA's) 'l
, ',\\ 'I.. _ _.
_.. _ __ ~.. -.. - _ -..__ _ -....._,..... . _.... _. A RSACTOR-OPERATOR Page 83 ' ANSWER: 080 (1.00) c.
[ + 1.'0 ] -
' REFERENCE: 1.
2203.038, " Primary To Secondary Tube Leakage", page 3.
2.
KA 000037A212 ( 3. 3 /4.1) 000037A212 ..(KA's) .., . ANSWER': 081 (1.00)
L b.
[+1.0) ! . , REFERENCE: , 1.
AA52003-008, " Steam Generator Tube Rupture", Objective-08.8, pages 19-20.
2.
KA 000038G011 (4.2/4.3) 000038G011 ..(KA's) ANSWER: 082 (1.00)
a.
[+1.0} REFERENCE: 1, 2202.006, " Loss of Feedwater", page 2.
_
2.- AA52003-010, " loss Of Feedwater", page 11.
.i 3'. KA 000054K304 (4.4/4.6) 000054K304 ..(KA's) p.
! ANSWER: 083.
(1. 00). . b.
. [ + 1. 0 ) - . . 'f i , ,
.-. - ~. > ! REACTOR' OPERATOR
- Page:84
- REFERENCE: - 1.
AA52002-007, " Electrical Distribution", page 32.
2.
KA 000058A203 (3.5/3.9) .000058A203 ..(KA's) ANSWER: 084-(1. 0 0 ) c.
[+1.0] ' REFERENCE: 1.
AA42102-002, " Pressurizer Pressure and Level Control System", Objective 5, pages 43-45.
-2.
KA 000028A212 (3.1/3.5) _ '"" l 000028A212 ..(KA's) ANSWER: 085 (1.00) a.
[+1.0) , REFERENCE: 1.
AA52003-011, " Loss of Offsite Power", Objective 11.6, page 21.
- .2.
KA 000056A288 (4.1/4.2)
000056A288 ..(KA's) -i ANSWER: 086 (1.00)
a -. [ +1'. 0] , - , u ,
. ... . . - - -. __ . .. . . .
F.
- REACTOR OPERATOR Page 85.
REFERENCE: 1.
2203.021, " Loss Of: Instrument Air", Attachment E, page 41.
2.
KA.000065A208 (2.9/3.3).
000065A208 ..(KA's) . ANSWER: 087 (1.00) a.
(+1.0] . REFERENCE: , 1.
1012.021, " Exposure Limits and Controls", page 3.
2.
KA 194001K103 (2.8/3.4) 194001K103 ..(KA's) ..;
RNSWER: 088 (1.00) d.
[+1.0] REFERENCE: -1.
1012.021, " Exposure-Limits and Controls", page 15.
'2.
KA 194001K103 (2.8/3.4) 194001K103 ..(KA's) ANSWER: 089 -(1.00) c.
(+1.0] ., > REFERENCE: 1.
'T/S BASES 3/4.4.7 . 2 '. KA 194001A114 (2.5/2.9) 194001A114 ..(KA's)
- 1 l
. . - . REACTOR' OPERATOR Page 86 )'; ANSWER: 090 (1.00) b.
[+1.0] REFERENCE: - -1.
1000.120, "ANO Fire Barrier Watch Program", page 7.
2.
KA 194001K116 (3.5/4.2) 194001K116 ..(KA's) , ANSWER: 091 (1.00) c.
[+1.0] REFERENCE: ! 1.
1000.128, " Industrial Safety and Occupational Health", page 27.
- 2.
KA 194001K109 (3.4/3.4) , , k 194001K109 ..(KA's) ANSWER: 092 (1.00) d.
[+1.01 REFERENCE: , 1.
1000.006, " Procedure Control", page 5.
2.
KA 194001A101 (3.3/3.4) 194001A101.. ..(KA's) e ANSWER: 093 (1.00) a.
[+1.0) - . . - - - -
- H., ' REACTOR OPERATOR Page~87 REFERENCE: 1.
1000.019, " Station' Security Requirements", page 13.
2.
KA 194001K105 '(3.1/3.4) ~ ~ -194001K105 ..(KA's) ,
ANSWER: 094 (1.00) d.
[+1.0) ., -REFERENCE:
1.. 1000.027, " Hold and Caution Card Control", page 39.
2.
KA 194001K102 (3.7/4.1)
194001K102 ..(KA's) , ANSWER: 095 (1.00)
-b.
[+1.0] , . REFERENCE: 1.
1015.001, " Conduct Of Operations, page 46.
2.
KA 194001K101 (3.6/3.7) 194001K101 ..(KA's) > - ANSWER: 096 (1.00)
be [+1.0) ' .:
- (
, REFERENCE: , f .1.
1015.011, " Conduct Of Operation"', page 51 ', 2.
}Ut 194001K101 (3.6/3.7) 194001K101 ..(KA's) , 'd a . j .; . - .
. . ... ._._-. . _.,... ... . . . . . REACTOR' OPERATOR' Page 88 , ! , . ANSWER: 097-(1.00)
b.
(+1.0) ' REFERENCE: 1.
1012.021, " Exposure Limits and Concrols", page 15.
l 2.
KA 194001K103 (2.8/3.4) 194001K103 ..(KA's) , ANSWER: 098 (1.00) j-b.
(+1.0) i
REFERENCE: ! 1.
2202.002, " Reactor Trip Recovery", page 2.
2.
KA 194001A102 (4.1/3.9)
194001A102 ..(KA's) i ANSWER: 099 (1.00) - d.
(+1.0) , REFERENCE: 1.
1015.028, " Operations Annunciator Control", page 3.
> 2.
KA 194001A109 (2.7/3.9)
194001A109- ..(KA's) ! .. . ANSWER:- 100 (1.00) b.
[+1.0)
.1 ! _ - ... -. , ,_, - i
. - - - -. . . .. . . . ., -...- REACTOR' OPERATOR Page 09 REFERENCE: 1.
1015.032, " OPS Procedure Users Guide", page 4.
2.
KA 194001A102 - (4.1/3,9) , 194001A102 ..(KA's) , P ' , V ' ,
(********** END OF. EXAMINATION **********)- - ,. . --. . -..
.. ,. . . . ~, - .. ..- . -. . -.,. . . 3 REACTOR OPERATOR Page
ANSWER KEY , ~ i l MULTIPLE CHOICE 023 c 001-c 024 d 002.
b 025 o 003 b 026 a , 004 d 027 a.
005 b 028 c 006 d 029 a , 007 a 030 a ! 008-a 031 a 009 a 032 d.
i 010 a 033 b a-011 a 034-d , 012 d 035 b- , 013 'b 036 c 014 c 037 .c
015 c 038 d 016 'd-039 c 017 c 040 a 018-b 041 a 019 c-1042 d i 020 c 043 d
021 d 044 b , 022 a 045 d
, -w 6,- es-., w, e,-... -* -- w
- - -
w .., ,, w-.. ,-
m.. -, -.. _. . _ _ _ -... _.. _ _. _ , . _ - . l ' REACTOR OPERATOR' Page 2' ANSWER KEY ' , 046 c ~ 069 d '047 b 070 a-048 d 071 b
r-049 b 072 d 050 b-073 a , 051 d 074 c 'i 052 a-075 c 053 b 076 a . 054 c 077 d
055 a 078 c l 056 c 079 a g 057 b 080 c
058 b 081 b- -059 a 082 a 060 d 083 b ) ~061, b 084 c ' 062 c 085 a 063-a 086-a
' -064 d 087' a 065 b 088 d- . 066 b 089 - c . 067-c-090 b 068-b 091 c
- )
H
- . REACTOR OPERATOR Page.- 3 ANSWER KEY I
-!
l 092 d , 093 a
094 d
- 095 b
096 h-097 b 098 b ci 099 d 100 b ,
L , .., ! t .
(********** END OF EXAMINATION **********) , i
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S. NUCLEAR REGULATORY COMMISSION SITE SPECIFIC EXAMINATION SENIOR OPERATOR LICENSE REGION
CANDIDATE'S NAME: FACILITY: Arkansas Nuclear One-2 REACTOR TYPE: .PWR-CE ' DATE ADMINISTERED:- 93/12/06 INSTRUCTIONS TO CANDIDATE: Use the answer sheets provided to document your answers.
Staple this cover sheet on top of the answer sheets.
Points for each question are indicated in parentheses after the question.
The passing grade requires a final grade of at least 80%. Examination papers will be picked up four (4) hours after the examination starts.
CANDIDATE'S TEST VALUE SCORE % 100.00 % TOTALS FINAL GRADE All work done on this examination is my own.
I have neither given nor received aid.
Candidate's Signature l
.-
__ _ __ -... _ _. ___., _ _ _.. _ _. . _. _ _ .-- . . .
- SENIOR' REACTOR OPERATOR Page
ANSWER-S H E.E T Multiple Choice (Circle or X your choice) ' If you. change your answer, write your selection in the blank, , , ? MULTIPLE CHOICE 023 a b 'c d '001-a b c d 024 a b c d ~
002 a b c d 025 a b c d
003 a b c d 026 a b c d-004 a b c d 027 a.
b .c d 005 a b c d 028 a b c d 006 a b c d 029 a b c d.
+ ' 007 a b c d 030 a b c d 008 a b c d 031 a b' c.
d . 009.
a b c d 032 a b-c d 010 a b c d 033 a b c d .011 a b c d 034 a b c d [ , 012-a b c d 035-a-b c' id 013 a.
b c d 036 a b ~c d 014 a b c d 037 a b c d.
, i 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 018-a b c d 041 a b c d 019 a b.
c d 042 a b.
c d t 020 a b c d 043 a b c d > 021.
a.
b c d-044 a b c d 022 a b c d 045 a b c d
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... . -.. . -- . - -.. .- . .... _ ~ . - ~.. - . .. ! ' -SENIOR REACTOR OPERATOR Page= 3L , ANSWER SHEET , Multiple Choice (Circle or X your choice) If.you change your answer, write your selection in the blank.
! '046 a b c d 069 a b c d ! 047 a b-c d 070 a b c .d ' 048 a b c d 071 a b-c d - 049 a b c d 072 a b c d ... 050~ a b c d .073 'a' .b-c d l b 051 a b c d 074 a b c d
__ 052 a b c d 075 a b c d 053 a b c d 076 a b c d' 054 a b c d 077 a b c d
i 055 a b c d 078 a.- b c' .d i 056 a.
b c d 079 'a b c d j 057 a b c.
d 080 a b c d .i 058 a' b c d 081 a b c d' 059 a b c d '082-a b c d
060 a b c d 083-a b c' d '!
061 a b c d 084 'a' b c d
i 062 a-b c d 085 a b' c d'
- i 063 a
b c 'd 086-a 'b c-d' - 064 a b c d 087-a b c d
J 065.
-a.
b c d-088-a .b c d .j 066 a b c d 089 a b c.
d
- !
i 067 a b c 'd 090 a b c d-l 068 a b c d 091 a b c d
l i . , . .. . -.. .-. ... -- -. - - - - .-
!- -SENIOR REACTOR' OPERATOR Page
ANGWER SHEET Multiple Choice (Circle or X your choice) If.you change your answer, write your selection in the blank.
'092 a b c d 093 a b c d .; '094 a b c d 095 a b c d ' 096 'a b 'c d , 097 a b c d . . 098 a b c d ' 099 a b c d ' 100 a b c d
! i f r , , !
, i !
(********** END OF EXAMINATION **********) l, v
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NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination the following rules apply: 1. Cheating on the examination means an automatic denial of your application and could result in more severe penalties.
2. After the examination has been completed, you must sign the statement on the cover sheet indicating that the work is your own and you have not received or given assistance in completing the examination.
This must be done after you complete the examination.
, 3. Restroom trips are to be limited and only one applicant at a time may.
' leave.
You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.
4.
Use black ink or dark pencil ONLY to facilitate legible reproductions.
l < 5.
Print your name in the blank provided in the upper right-hand corner'of the examination cover sheet and each answer sheet.
. 6. Mark your answers on the answer sheet provided.
USE ONLY THE PAPER PROVIDED AND DO NOT WRITE ON THE BACK SIDE OF THE PAGE.
7.
Before you turn in your examination, consecutively number each answer sheet, including any additional pages inserted when writing your answers on the examination question page.
8. Use abbreviations only if they are commonly used in facility literature.
Avoid using symbols such as < or > signs to avoid a simple transposition error resulting in an incorrect answer.
Write it out.
J 9. The point value for each question is indicated in parentheses after the question.
- 10. Show all calculations, methods, or assumptions used to obtain an answer to- - any short answer questions.
, 11. Partial credit may be given except on multiple choice questions.
Therefore, . ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER-BLANK.
12. Proportional grading will be applied.
Any' additional wrong information that is provided may. count against you.
For example, if a question is' worth one point and asks for four responses, each of which is worth 0.25 points, and you give five responses, each of your responses will be worth 0.20 points.
If one of your five responses is incorrect, 0.20 will be deducted and your total. credit for that question will be 0.80 instead of 1.00 even though you got~the four correct answers.
13. If the intent of a-question is unclear, ask questions of the examiner only.
. -, -- - - - - . - - -- - . - _ - ,. - - . -. _
_..., . . _ _ _. . -.. _. ._.
._ _ _ _ _ f Page
14. When turning in your examination, assemble the completed examination with examination questions, examination aids and answer sheets.
In addition, turn in all scrap paper.
15. Ensure all information you-wish to have evaluated as part of your answer is on your answer sheet.
Scrap paper will be disposed of immediately following the examination.
16. To pass the examination, you must achieve a grade of 80% or greater.
17. There is a time limit of four (4) hours for completion of the examination.
- 18. When you are done and have turned in your examination, leave the examination area (EXAMINER WILL DEFINE THE AREA).
If you are found in this area while the_ examination is still in progress, your license may be denied or revoked.
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- QUESTION: 001 (1.00)
- The following plant conditions exist: 1.
Unit 2 has tripped from 100% power due to a loss of offsite power.
2.
Emergency Diesel Generator 2K4A has failed to start.
i 3.
RCS pressure is 2230 psia.
4.
Two (2).CEAs have failed to fully insert into the core.
5.
RWT level is 70%. 6.
The Boric Acid flow transmitter down stream of the Boric Acid' Makeup ' pumps has just been discovered to be plugged.
7.
Reactor operators are performing Emergency Operating Procedure-2202.001, " Standard Post Trip Actions".
WHICH ONE (1) of the following mechods is available for emergency boration given the above conditions? a.
HPSI pump 2P-89A through High Pressure Injection valves.
b.
Gravity feed from RWT through 2CV-4950-2 to the Charging pump 2P36C.
c.
Normal Emergency Boration flowpath through 2CV-4916-2 to the Charging pump 2P36B.
d.
Gravity feed through 2CV 4921-1 to the Charging pump 2P36A.
QUESTION: 002 (1.00) WHICH of the following actions should bo performed if the Thot input to the Reactor Regulating System (RRS) fails HIGH? E a.
Commence boration.
b.
Take letdown control to MANUAL.
c.
Commence dilution.
d.
Take backup heaters to OFF.
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.- SENIOR REACTOR OPERATOR.
Page
' t i QUESTION: 003 (1.00) WHICH ONE (1) of the following describes the CEDMCS Withdrawal Prohibit (CWP) alarm function? a.
Initiated by the CEA Position Indication System in response to a PRETRIP condition on DNBR and prohibits ALL movement of regulating , CEAs.
' b.
Initiated by the Plant Protection System (PPS) in response to a PRETRIP condition on DNBR and prohibits ALL OUTWARD group movement of CEAs.
c.
Initiated by the CEA Position Indication System of the Plant Computer in response to OUT OF SEQUENCE / OVERLAP condition and
prohibits ALL OUTWARD movement of CEAs.
. d.
Initiated by the Plant Protection System (PPS) in response to OUT OF SEQUENCE / OVERLAP condition and prohibits ALL movement of CEAs.
!
' QUESTION: 004 (1.00) i WHICH ONE (1) of the following conditions exists if the RED indicating light for the RCP 2P32D is NOT lit? (Assume the bulb has been verified as good).
' a.
The 2P32D RCP trip circuit has lost continuity and the ONLY means of tripping the pump is to open the breaker locally.
b.
Thel 2P32D RCP trip circuit has lost continuity and the local handswitch must be used to trip the pump.
c.
The 2P32D RCP start circuit has lost continuity and the ONLY means of starting the pump is to close the breaker locally.
d.
The 2P32D RCP start circuit has lost continuity and the local handswitch must be used to start the pump.
'I _ _ __ _ . . - _ _ ._
. .. .. _ - . _ _ __ _ __ _ _. _ - - - . __ . > SENIOR REACTOR OPERATOR Page: 9-I i QUESTION: 005 (1.00) WHICH ONE (1) of the following is the basis for NOT starting the fourth-J Reactor Coolant Pump for Unit 2 until RCS temperature is greater than 500
degrees F.? !, a.
To prevent exceeding RCS heatup rate limits.
, r b.
To minimize excessive RCP starting currents.
> c.
To limit Steam Generator tube stresses.
, d.
To limit core uplift.
, , ) QUESTION: 006 (1.00) ., WHICH ONE (1) of the following is actuated by LOW Pressurizer pressure. signal?
a.
Containment Spray System (CSS) ' b.
Penetration-Room Ventilation System (PRVS) c.
Main Steam Isolation System (MSIS) d.
Containment Cooling System (CCS) )
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. .. - ._..- .. _. ~. . -... - .. .. _. . . . _ _ SENIOR REACTOR OPERATOR Page 10.
QUESTION: 007 (1.00) .WHICH ONE (1) of the following is the reason thirty (30) degrees F.
is chosen as the setpoint for the Margin To Saturation (IfDS) alarm?- a.
To ensure net positive suction pressure (NPSH) is maintained for the reactor coolant pumps (RCPs).
b.
To ensure that RCS voiding is prevented.
' ' c.
To ensure that actual subcooling margin is maintained due to instrument errors.
d.
To ensure that actual subcooling margin is maintained when the'RCPs are stopped.
QUESTION: 008 (1.00) WHICH ONE (1) of the following systems provides cooling to the~ Containment Cooling System following a design basis Loss Of Coolant Accident (LOCA)? a.
Chilled Water b.
Component Cooling Water ) c.
Service Water i d.
Auxiliary Cooling Water i l
,
. V ' SENIOR' REACTOR OPERATOR Page 11 ' QUESTION: 009 (1.00) .VGIICH ONE (1) of the following conditions will prevent starting. Heater Drain ' Pump 2P8A at 40% reactor power? a.
A Main Feed Pump running and 2CV0742 open.
' b.
.A Main Feed Pump running and 2CV0742 closed.
c.
B Main Feed Pump running'and 2CV0742 open.
d.
B Main Feed Pump ' running and 2CV0742 ' closed.
, -i ' QUESTION: 010 (1.00) WHICH ONE (1) of the following conditions will AUTOMATICALLY close Feedwater , to Steam Generator Isolation valve 2CV-1024-1? a.
LOW level in Steam Generator "A" b.
HIGH level in Steam Generator "A" c.
MSIS d.
CIAS , h
't ,
,9-g y3 ,y = ,. - =
_.. , -SENIOR REACTOR OPERATOR Page 12 , QUESTION: 011 (1.00) The'following. condition exists: 1.
Unit 2 is at 5% power.
, 2.
EFAS has actuated.
3.
EFW pump suction is aligned to the Startup and Blowdown Demineralizer effluent.
r 4.
Suction pressure for the Emergency Feedwater (EFW) pumps is'10 psigt and DECREASING.
WHICH ONE (1)-of the following automatic actions will occurfif EFW pump suction pressure continues to DECREASE to 5 psig? a.
2CV-0789-1 and 2CV-0795-2 will automatically OPEN to supply _ water from the Condensate Storage Tank.
b.
2CV-0716-1 and 2CV-O ril-2 will automatically OPEN supply water f rom the Service Water System.
c.
2EFW-0706 will automatically OPEN to supply water.from the Condensate Storage Tank.
d.
2EFW-16, Condensate Storage Tank Suction Bypass valve, will automatically OPEN to supply additional water from the Condensate Storage Tank.
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SENIOR REACTOR OPERATOR Page 13 , QUESTION: 012 (1. 0 0 ). The following conditions exist: Steam Generator "A" Steam Generator "B"
1.
Level 25% & decreasing 20% & increasing 2.
Steam Pressure 725 psig '825 psig 3.
Feedwater Pressure 1200 psig 1200 psig.
WHICH ONE (1) of the following is the current status.of the Emergency ' Feedwater System (EFS) given the above plant conditions? Assume all plant systems respond as designed.
- , a.
EFAS has not ocu red but will AUTOMATICALLY actuate when Steam ' Generator "A" level decreases to 23%. , b.
An EFAS trip condition exists and EFW is being supplied to BOTH steam generators.
c.
An EFAS trip condition exists and EFW is being supplied-to.ONLY Steam Generator "B".
d.
An EFAS trip' condition exists and EFW is BLOCKED to BOTH steam generators.
> > ! I i i n m = ., - , .. SENIOR REACTOR' OPERATOR Page 14 OUESTION: 013 (1.00)
- The following conditions exist:
1.
Unit 2 is in HOT STANDBY.
2.
EFW is-in MANUAL control.
3.
EFW control valves '2CV-1025-1 and 2CV-1075-1 are throttled to BOTH Steam Generators.
4.
Simultaneous EFAS A#1 and MSIS #1 actuation signals are received due to an I & C technician error.
WHICH ONE (1) of the following is the affect of this-error on EFW control valve 2CV-1025-1 and 2CV-1075-1 operation? .a.
Valves remain throttled because they are in MANUAL control.
b.
Valves remain throttled because EFAS A#1 and MSIS #1 have NO affect' on these valves.
c.
Valves go FULL open due to the EFAS A#1 signal.
d.
Valves go FULL closed due to the MSIS #1 signal.
QUESTION: 014 (1.00) WHICH ONE (1) of the following conditions will result in AUTOMATICALLY closing Regenerative Waste Discharge valve 2CV-4424 (2T-92 discharge to flume) ? a.
Tank inlet valve open.
b.
Discharge line pH reading of 8.3.
c.
Low instrument air pressure of 50 psig.
d.
Waste Inlet Radiation Monitor 2RE-4447 HIGH alarm trip.
.
.. . . - -- . - -.. -. ,, ' 'i , . SENIOR REACTOR OPERATOR Page 15 t-QUESTION: 015 (1.00) l WHICH ONE (1) of the following discharges to the Drain Collection Header.
(DCH) ? a.
Blowdown tank drains.
b.
HPSI relief valves.
l c.
LPSI relief valves.
. d.
Containment sump drains, i .
QUESTION: 016 (1.00) WHICH ONE (1) of the following sets of Waste Gas Tank parameters is acceptable? (2104.010, " Unit 2 Waste Gas Analyzer Operation", Attachment: "A" is attached.)
' pressure Hydrogen Oxygen a.
200 psig 40%- 12% -, b.
200 psig 20% 5% c.
250 psig 60% 3% ! d.
250 psig 6% 10% .
- ) l
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i ' ~ l H SEMIOR REACTOR OPERATOR LPage 16' I i . QUESTION: 017 (1.00) WHICH ONE (1) of.the.following is the cause for an AREA Radiation Monitor to Lemit a' LOW beeping sound?- , a.
Monitor is.placed in test.
b.
. Monitor has-failed LOW.
c.
A valid HI radiation alarm.
d.
Battery backup is discharged, t QUESTION: 018 (1.00) WHICH ONE (1) of the following inputs is provided by Loop 2 hot. leg temperature instrumentation? a.
ESFAS bypass permissive for steam generator high/ low water level trip signals .; b.
AWP (Automatic Withdrawal Prohibit) in CEDMCS c.
LTOP isolation valve misalignment annunciation when valves are'OPEN with temperature less than 240. degrees F.
d.
Steam Dump and Bypass. Control System Modulating Opening Signal-
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- i SENIOR REACTOR OPERATOR Page'17'
,
QUESTION: 019 (1.00) The following conditions exist: 1.
The plant is operating at 75% power and the latest leak rate data indicates: , .
11.5-GPM - Total RCS leakage rate
2.5 GPM - Leakage into the Quench Tank
2.2 GPM - Leakage into the Reactor Drain Tank
3.5 GPM - Leakage past check valves from RCS to SI system
1.2 GPM - Total primary to secondary leakage (Assume distributed over all S/Gs) WHICH ONE (1) of the following is the current RCS leakage status per Technical Specifications 3.4.6.2, " Reactor Coolant System Leakage"? - a.
PRESSURE BOUNDARY LEAKAGE that requires shutdown.
b. UNIDENTIFIED LEAKAGE that requires shutdown.
c. IDENTIFIED LEAKAGE that does NOT require shutdown.
d. PRIMARY to SECONDARY LEAKAGE that does NOT require shutdown.
QUESTION: 020 (1.00) T To WHICH ONE (1) of the following locations are the Safety Injection. Tanks (SIT) vented when lowering pressure in Mode 3? a.
Containment atmosphere b.
Reactor Drain Tank < c.
Quench Tank d.
Holdup Tank
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..- .... . ... -- . _ . - - . ..~ . SENIOR REACTOR OPERATOR-Page 18 ) QUESTION: 021, (1.00) , The following conditions exist: 1.
Unit 2 has tripped due to a small break loss of coolant accident > (LOCA).
i 2.
The event is ongoing and the Recirculation Actuation System (RAS) has initiated.
3.
RCS pressure has increased to 1500 psia.
4.
BOTH High Head Safety Injection (HPSI) pumps continue to operate at: full capacity.
WHICH ONE (1) of the following provides cooling for the HPSI pumps? a.
Flow through the pumps to the RCS is sufficient to cool the pumps under these conditions.
b.
Mini-Recirc flow to the RWT is sufficient to cool the-pumps under these conditions.
c.
Mini-Mini-Recirc flow to the pump suction is sufficient to cool the pumps for up to thirty eight (38) minutes.
d.
Mini-Recirc flow to the containment sump is sufficient to' cool the pumps for up to thirty eight (38) minutes.
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.__ i SENIOR' REACTOR OPERATOR-Page 19 I , ' QUESTION: 022 (1.00) The following p1' ant conditions exist: 1.
RCS pressure.is 2250 pala.
2.
Pressurizer Pressure Control System (PPCS) setpoint is 2250. psia.
3.
RCS boron' concentration is 500 ppm boron.
4.
Pressurizer boron concentration is-350 ppm boron.
-WHICH ONE (1) of the following is the correct method to equalize RCS boron concentration per 2103.005, " Pressurizer Operation"? a.
ENERGIZE all Pressurizer Backup Heaters and place-"A" RCP spray.
_ valve in MANUAL and OPEN to maintain RCS pressure within + 10 psia of desired setpoint.
, b.
ENERGIZE all Pressurizer Proportional Heaters and place "A" RCP spray valve'in MANUAL and OPEN to maintain RCS pressure within +'10 psia of desired setpoint.
RAISE Pressurizer-Pressure Control System setpoint to 2300Lpsia AND' c.
, energize Pressurizer Proportional Heaters to maintain Pressurizer Spray flow.
d.
DECREASE Pressurizer Pressure Control System setpoint to 2220 psia AND energize Pressurizer Backup Heaters to maintain Pressurizer Spray flow.
QUESTION: 023 (1.00) , Pressurizer level indication on the Remote Shutdown Panel may differ from that of Level Channel 1 as indicated on Panel 2C04.
This is because'the Remote Shutdown Panel receives its signal from' ? a.
Level Channel 2 and is NOT temperature compensated.
. b.
Level Channel 2 and is temperature compensated.
c.
SPDS which is temperature compensated.
d.
SPDS which-is NOT temperature compensated.
. !
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. . . . -. . .. .. - -. - - - -SENIOR REACTOR OPERATOR Page 20 , . QUESTION: 024 (1.00) , The following plant conditions exist: 1.
Unit 2 is at 100% power.
2.
RCS pressure is 2405 psia.
3.
The Reactor Protection System (RPS) has NOT actuated.
4.
All Reactor Protection System setpoints have been calibrated to plant design specifications.
WHICH ONE (1) of the following results from RCS pressure INCREASING to 2450 psia? Assume no additional failures occur.
i a.
The HIGH Pressurizer pressure setpoint of the RPS is reached and Reactor Trip Breakers will OPEN to DE-ENERGIZE the control element assembly (CEA) holding coils.
b.
-The HIGH Pressurizer pressure setpoint of the Diverse Scram System (DSS) is reached DE-ENERGIZING the trip coil to open a contact interrupting output power from both Motor Generator (:MG) sets.
c.
The HIGH Pressurizer pressure setpoint of the RPS is reached and the MG set 480 VAC output breakers will OPEN to DE-ENERGIZE the control > element assembly (CEA) holding coils, d.
The HIGH Pressurizer pressure setpoint of the Diverse Scram System (DSS) is reached ENERGIZING the trip coil to open a contact r interrupting output power from both Motor Generator (MG) sets.
QUESTION: 025 (1.00) WHICH ONE (1) of the following is the reason for a FLASHING indication on the
' Saturation Margin Monitor (SMM) ? I a.
Indicated value is below the saturation margin alarm setpoint.
b.
Indicated value is below 30 degrees F saturation margin.
c.
Indicated value is erroneous.
i d.
Indicated value is superheated.
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- _ , SENIOR REACTOR-OPERATOR Page 21 .; . QUESTIONi 026 (1.00) l WHICH ONE (1) of the following will result in the loss of Containment Spray i Pump 2P-35A?' i a.
Loss of 6900 VAC Buses 2Hl.
.
b.
Loss of 6900 VAC Buses 2H2.
t c.
Loss of 4160 VAC Buses 2A3.
, d.
Loss of 4160 VAC Buses 2A4.
+ QUESTION: 027 (1.00) . WHICH ONE (1) of the following is the MINIMUM condition that will INITIATE- . actuation of the Containment Spray System? a.
CIAS actuation.
l b.
CIAS and SIAS actuation.
c.
Containment pressure at 24.0 psia and CIAS actuation.
d.
Containment pressure at 24.0 psia and SIAS actuation.
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- QUESTION: 028_
(1.00) ~ WHICH ONE (1) of the following AUTOMATIC actions-take place following a trip of the' Containment Purge Exhaust fan to prevent:overpressurization of the containment? a.
Exhaust dampers go 100% open to relieve pressure.
i b.
Supply dampers throttle to control pressure.
c.
Supply fan trips after a ten (10) second time delay.
' , d.
Purge isolation valve 2CV-8284 closes.
QUESTION: 029 (1.00) WHICH ONE (1) of the following conditions could prevent the Spent Fuel [ Pool.
Cooling System from maintaining pool temperature below the' design-limit.of 150 degrees F.? a.
Component Cooling water.(CCW) flow to the Spent. Fuel Pool' heat exchanger is 1500 gpm.
b.
The reactor has been shutdown for 120 hours when spent fueluis transferred to the Spent Fuel Pool.
c.
Only one (1) Spent Fuel Pool pump'is OPERABLE.
d.
Spent Fuel Pool level is reduced to 410 feet 2 inches.
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.. - .- . . - ' SENIOR-REACTOR OPERATOR Page 23 ~ QUESTION: 030 (1.00) The following conditions exist: 1.
Unit 2 has. tripped from 100% power one (1) minute ago.
2.
Steam Generator "A" level is 95%. 3.
Steam Generator "B" level is 40%. 4.
High Level Override Level Selector switch is selected to the BOTH.
position.
WHICH ONE (1) of the following is the expected status of the Feedwater Control' System (FWCS)? (Assume NO operator action.)
a.
Feedwater is isolated to Steam Generator "A" and at St~ flow demand to Steam Generator "B".
b.
Feedwater is isolated to BOTH Steam Generators "A" and "B".
c.
Feedwater is at 5% flow demand to BOTH Steam Generators "A" and "B".
d.
Feedwater is at 5% flow demand to Steam Generator "A" and at 100% flow demand to "B" Steam Generator.
QUESTION: 031 (1.00) The following-conditions exist: 1.
Unit 2 is operating at 100% power.
, 2.
Steam Generator Level selector switch is selected to BOTH.
' WHICH ONE (1) of the following will cause the "B" Steam Generator Flow Demand Signal to DECREASE? a.
"B" Steam Generator Level transmitter fails HIGH.
l ' b.
"B" Steam Generator Steam Flow transmitter fails HIGH, c.
"B" Steam Generator Feedwater Flow transmitter fails LOW.
d.
"B" Steam Generator Pressure transmitter-fails LO.. -..-. , , - _ - -. _ _ - ~ -.... ... -.. - -~ ._ , SENIOR REACTOR OPERATOR Page 24
QUESTION: 032 (1.00) .; ' The following conditions exist: 1.
Unit 2 is operating at.100% power.
2.
Operators are performing:the *MSIV Quarterly Partial Stroke Test".
WHICH ONE (1) of the following will result from holding the EXERCISE switch in the CLOSE position for one (1) minute? a.
The valve will stroke closed until the. GREEN closed light is , illuminated, then re-open automatically.
b.
The valve will stroke closed until'it is 10% closed, then re-open automatically.
c.
The valve will stroke closed until it is 10% closed, then remain in
that position until re-opened manually, , d.
The valve will continue to stroke closed until the EXERCISE switch is placed in the OPEN position.
QUESTION: 033 (1.00) WHICH ONE (1) of the following is NOT an AUTOMATIC action which'results from a HIGH alarm trip of Unit 2 Control Room Ventilation Radiation Monitor 2RE-8750- ' 1? a.
Unit 1 Control Room Isolation dampers close.
l b.
Unit 1 Supply fan VSF-8 stops.
c.
Control Room Emergency Supply fan 2VSF-9 starts.
d.
Control Room R.ecirculation dampers open.
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.. _. _ .. .. - . - _ _ ~ _ _. _. _.. . _..._._. - . SENIOR. REACTOR OPERATOR .Page 25 ! QUESTION: 034 (1.00) The~following conditions exist: ' 1.
Unit 2 is in HOT SHUTDOWN.
2.
Operators are placing SDC in service.
3.
SDC Heat' Exchanger Cross-tie valves 2SI-5A and 2SI-5B are OPEN.
4.
Containment Spray Header MOVs 2CV-5612-1/2CV-5613-2 are OPEN for testing.
WHICH ONE (1) of the following will occur when LPSI pump 2P60A is started? ' a.
BOTH 2CV-5612-1 AND 2CV-5613-2 will close AUTOMATICALLY.
b.
2CV-5612-1 will close AUTOMATICALLY but 2CV-5613-2 will remain open.
'i c.
LPSI Header Isolation valve 2SI-6 will close AUTOMATICALLY but 2SI-5B will remain open.
, d.
2SI-5A will close AUTOMATICALLY.
! QUESTION: 035 (1.00) ' WHICH ONE (1) of the following is a possible source of inventory for the
Quench Tank? a.
Reactor Drain Tank discharge b.
SIT drain lines . c.
Service Water System makeup d.
RCP Controlled Bleedoff i
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SENIOR' REACTOR OPERATOR Page 26~ 's
QUESTION: 036 (1.00) The following conditions exist: 1.
Unit.2 has tripped due to a large steam break inside containment.
' ' 2.
CSAS has been actuated for one (1) minute.
- 3.
Sodium Hydroxide pump 2P136A has NOT started.
, ' .WHICH ONE (1) of the following could be a reason the-2P136A pump is NOT running? (Assume all systems are performing as designed.)
a.
The pump start time delay has-not timed out.
b.
Sodium Hydroxide pump 2P136A discharge valve is closed.
c.
Sodium Hydroxide tank outlet valve is only-75% open.
> d.
Sodium Hydroxide tank. level is 50%. QUESTION: 037 (1.00) Hydrogen Analyzers must be placed in service within fifteen (15) minutes 'following a Loss.Of Coolant Accident'(LOCA)? This is based on NUREG 1.97.
.' which-requires a hydrogen sample within .? a.
thirty (30) minutes of a LOCA and it takes fifteen (15) minutes for the analyzers to warmup.
- b.
thirty (30) minutes of a'LOCA and it takes approximately 11 minutes to transport the gas to the analyzers.
c.
one (1) hour of a LOCA and it takes forty five (45) minutes for the.. analyzers to warmup, d.
one (1) hour of a LOCA. and it takes thirty (30). minutes for' the analyzers to warmup and analyze a sample and approximately 11 minutes to transport the gas to the analyzers.
. s.
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>. . . . . _ SENIOR REACTOR. OPERATOR Page. ; 27-QUESTION: 038 (1.00) WHICH ONE (1) of the following conditions will PREVENT movementfof the Refueling Machine?- a.
Spreader is NOT extended.
b.
Fuel assembly is in the full up position.
c.
Suspended load is at setpoint.
'd.
Fuel Hoist is operating.
- QUESTION: 039 (1.00) WHICH ONE (1) of the following conditions renders the Service Water System IMOPERABLE per Technical Specification 3.7.3, " Service Water System"?
The Red ESF bus supplies power to both' Service Water pump 2P4B and' a.
' the Auxiliary Cooling. Water System-(ACW).
b.
Service Water pump discharge filter. delta pressure-is 11 paid.
c.
The Green ESF bus supplies power to both the Service-Water pump 2P4B and the 2P4B sluice gate.
d.
Service Water pump suction pressure is 11'psig.
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i-l SENIOR _ REACTOR OPERATOR Page 28: ^l .. QUESTION:. 040 (1.00)
. j A' major Turbine' Building _ Instrument Air. header rupture has occurred on_ Unit 2.-
WHICH ONE (1) of the following is expected to provide-Instrument Air to Unit 2 from Unit 1?
a.
Unit 1 Crosstie valve 2SV-3015 AUTOMATICALLY opens.
! b.
Unit 1 Crosstie valve 2CV-3004 AUTOMATICALLY closes.
, c.
MANUALLY open IA TO UNIT 1 CROSS CONNECT valve IA-51.
d.
MANUALLY open Unit 1 to Unit 2 Crosstie valves 2CV-3015 and 2CV- , 3004.
!
QUESTION: 041 (1.00) , J WHICH ONE (1) of the following describes the reason for the Transient CEA ! Insertion Limits? ' a.
Ensures a MAXIMUM shutdown margin assuming an excess' steam demand accident at the beginning of core life.
t b.
Ensures a MAXIMUM shutdown margin assuming the failure of'the highest reactivity worth single CEA.
c.
Ensures axial peaking factors are within acceptable levels.
d.
Ensures the potential effects of'a CEA ejection accident are limited ' to acceptable levels.
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QUESTION: 042 (1.00) The following Reactor Coolant Pump conditions exist: "A" RCP "B" RCP 1.
RCP Vapor Seal Pressure 1600 psia 1250 psia 2.
Pump Shaft Vibration 12 mils 14 mils 3.
Pump Frame Vibration 2 mils 1 mil 4.
RCP Controlled Bleedoff 1 gpm 4 gpm WHICH ONE (1) of the following actions is. required per 2203.025, "RCP ,
-Emergencies"?- ASSUME the reactor has been tripped. ~ , a.
Secure "A" RCP due to HIGH RCP' Vapor Seal Pressure.
I b.
Secure "B" RCP due to HIGH Pump Shaft Vibration.
~ c.
Secure "A" RCP due to.HIGH Pump Frame Vibration.
, d.
Secure "B" RCP~due to'HIGH Controlled Bleedoff flow.
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.. . ,, _ - _ _ _ - -. _ ... ... _. _ _ _ _.. .. - -. l SENIOR REACTOR OPERATOR Page.30-QUESTION:.043 (1.004
The~following conditions exist: 1.
Unit 2 is at 100% power.
2.
Total Component Cooling Water flow is 250 gpm to RCP 2P32A.
3.
RCP 2P32A upper. seal cavity pressure is 1125 psia.
4.
RCP 2P32A middle seal cavity pressure is'1125 psia.
5.
Pressurizer pressure is 2250 psia.
WHICH ONE (1) of.the following describes the condition of RCP 2P32A?-
a.
Low Component Cooling : Water (CCW) flow exists.
b.
Upper seal failure has occurred.
c.
Middle seal failure has occurred.
d.
Lower seal failure has occurred.
! , b
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- l - SENIOR REACTOR OPERATOR Page 31
QUESTION: 044 (1.00) { The following conditions exist: , 1.
Unit 2 has tripped from 100% power and reactor power is DECREASING.
+ 2.
Emergency Boration was initiated at.0800 hours when CEAs #40 and #41 failed to fully insert.
3.
A Pressurizer' level control malfunction has resulted in a Pressurizer level of 95% and INCREASING.
i 4.
Tavg is 565 degrees F. and DECREASING slowly.
5.
Pressurizer pressure is 2400 psia and INCREASING.
. , 6.
Operators are SLOWLY inserting CEA #40 manually and it will'be fully { inserted at 0815 hours.
7.
CEA #41 remains fully withdrawn and immovable.
8.
Operators desire to terminate RCS makeup to prevent RCS , overpressurization.
' i WHICH ONE (1) of the following conditions will allow termination of RCS makeup? < a.
CEA #40 is fully inserted.
, b.
Thirty (30) minutes of Emergency Boration has been completed, c.
Pressurizer level reaches 100%. t d.
Pressurizer pressure reaches 2450 psia.
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-- . .. - -_...-. - . . . - - _ -.__ . . ~ SENIOR REACTOR' OPERATOR Page 32 i QUESTION: 045 (1.00) WHICH ONE (1) of the following is the basis'for'the limit on boron concentration and temperature in the RWT according to Technical Specification: , 3.5.4, " Refueling Water Tank"? j a.
Prevents corrosive attack of system piping during maximum flow conditions such as a boration event.
b.
Ensures integrity of CVCS pump suction piping by presanting thermalj shock conditions during a boration event.
c.
Prevents excessive motor current for the Charging pumps during a maximum flow condition such as a'boration event.
d.
Ensures reactor will remain subcritical in the cold condition I following mixing of the RWT and RCS water volumes during a boration' event.
, . -QUESTION: 046 (1. 0 0 ) - The following conditions exist: ' , 1.
2P33A is is in PTL for maintenance.
2.
2P33B is supplying loop I CCW.
3, 2P33C is supplying loop II CCW.
4.
A rupture occurs in loop II CCW, causing pressure to decrease to less than 80 psig.
WHICH ONE (1) of the following AUTOMATIC actions will result from the above conditions? a.
CCW Loop I Crossover valves 2CV-5220 and 2CV-5230 close.
i b.
CCW Loop II Crossover valves 2CV-5221 and 2CV-5232 close.
~ c.
CCW Loop II Crossover valves 2CV-5221 and 2CV 4232 open.
< d.
An IMMEDIATE reactor trip on loss of CCW flow to the Reactor Coolant pumps.
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_ . i t . SENIOR REACTOR OPERATOR Page 33 , t-QUESTION: 047 (1.00) , The following conditions exist: - e 1.
Unit 2 is at 100% power.
l 2.
Pressurizer spray valve 2CV-4651 has failed to close on' demand.
' i 3.
Operators are taking actions per 2203.028, "PZR Systems-Malfunctions", Step 1.
WHICH ONE (1) of the following is the reason the control switch ~for 2CV-4651 , .is placed to OPEN for one (1) second then to CLOSE for five (5)' seconds? i a.
This action OVERRIDES the Master Controller and allows the valve to be closed MANUALLY.
I b.
This action activates the mechanical hammer mechanism which assists ) in seating the valve.
c.
This action activates the QUICK CLOSE circuitry for the valve.
d.
This action opens additional CLOSE solenoids which provide greater air pressure to seat the valve.
.,
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- . - - .. .-.. ~ .- SENIOR'REACTORLOPERATOR Page 34-QUESTION: 048 (1.00) The following conditions exist: 1.
A steam line break exists upstream of the "A" Steam Generator MSIV-on Unit 2, 2.
MSIS has automatically initiated.
, WHICH ONE (1) of the following conditions could result if a steaming flowpath ' from the unaffected steam generator is NOT established immediately following dryout of the affected steam generator? a.
Rapid repressurization of the RCS and subsequent Pressurized. Thermal Shock (PTS) conditions.
b.
Inability to open "B" S/G MSIV due to pressure' difference created when affected steam generator reaches dryout conditions.
c.
A rapid INCREASE in T-cold of the UNAFFECTED loop resulting in an interruption of natural circulation.
i d.
A INCREASE in the core exit temperatures resulting in an interruption of natural circulation.
QUESTION: 049 (1.00) WHICH ONE (1) of the following discriminates between a steam line break inside containment and a small break LOCA? a.
Containment pressure b.
RCS pressure c.
Containment temperature d.
RCS temperature , . - -
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i QUESTION: 050 (1.00) The.following conditions exist: , 1.
1brbine roll is in progress.
2.
1800 rpm is selected.
3.
Turbine speed is 1700 rpm and increasing rapidly.
- WHICII ONE (1) of the following actions should be performed? a.
Select 800 rpm on-the speed set-rpm selector, b.
Select Close Valves on the speed set-rpm selector, c.
Break vacuum and verify turbine is tripped.
d.
Trip the reactor and verify the turbine is ti:1pped.
, t . QUESTION: 051 (1.00) TGIICH ONE (1) of.the'following is an indication that natural circulation exists following a '.oss of electrical power? (All temperatures are degrees JP. ) RCS Subcooling Thot Tcold CETs
a.
550 525 540
, b.
560 500 570 ' c.
540 500 530 d.
540 490 560'
) s
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SENIOR REACTOR OPERATOR-Page-36
't .. QUESTION: 052 (1.00) The.following conditions exist: ' 1.
Unit 2 has tripped from 100% power due to a. Loss of Offsite power.
2.
Diesel generators are running with NO' Service Water (SW) cooling.
t .WHICH ONE (1) of the following is the MAXIMUM time period that-the diesels can ! ' run with NO SW cooling.WITHOUT suffering damage? a.
Three (3) minutes ! o b.
Ten (10) minutes , c.
Fifteen (15) minutes d.
Diesels can run indeflyiral.y since they provide their own cooling
flow.
' ' i i 1-QUESTION: 053 (1.00) ' The following conditions exist: , 1.
The control room operator has received a report of'a small-fire in-l the 317' elevation of the Auxiliary Building.
. . 2.
The individual reports the fire is contained in the corner.of the . 2T20B room where oily rags have been stored incorrectly.
! 3.
The same individual reports there is no. smoke, the area is well - ventilated, and he has an ABC Dry Chemical Extinguisher.
4.
The individual has requested permission to extinguish theLfire.
E WHICH ONE (1) of the following conditions prevent this authorization? a.
The fire is caused by a hazardous chemical.
! I b.
The potential for smoke exists and the individual has no respiratoryf ' equipment.
j c.
The extinguisher is incorrect for this type of fire.
! d.
The fire is in a radiological control area (RCA).
' ! i ,
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. ..- . ~- . -. -. -. - - ,_ - ' SENIOR' REACTOR' OPERATOR 'Page 37~ , , QUESTION: 054 (1.00) a-WHICH ONE (1) of the following is an indication that is required by. Technical' Specifications to be displayed at the Remote. Shutdown Panel (2C80) ? { a. Reactor Trip Breaker Indication b. Steam Generator Pressure c. RCS Subcooled/Superheat Monitor . d. Refueling Water Tank Level j , , QUESTION: 055 (1.00) i '
WHICH ONE (1) of the following operator actions is required during an , evacuation of the Unit 2 control room due to a fire in the cable spreading room per 2203.014, " Alternate Shutdown"?
a.
Trip two (2) Reactor Coolant pumps (one in each loop) and leave two l (2) running.
b.
Trip ALL f our (4) ' Reactor Coolant pumps.
l ~ c.
Trip the "B" Loop Reactor Coolant pumps.
d.
Leave ALL R? actor Coolant' pumps running if seal water can be established within five (5) minutes.
. . QUESTION: 056 (1.00) .; WHICH ONE (1) of the following conditions. represents a LOSS of Unit 2 i Containment Integrity? Unit 2 is in REFUELING Mode, a.
Equipment Hatch closed with ONLY four (4) bolts., d
b.
One (1) Air Lock door is OPEN.
, c.
Containment Purge is in progress.
_ d.
An electrical penetration's seals are removed for repairs.
! .
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- -... _ -_.._ _ _ ._, Y , ISENIOR REACTOR OPERATOR Page 38-e i i i QUESTION: 057 (1.00)- .! '! The following conditions exist: 1.
Unit 2 has tripped due to a loss of Feedwater.
-f 2.
Operators have entered Functional Recovery Guidelines.
WHICH ONE (1) of the following is the criteria for going to HR-4, "Once
Through Cooling" to provide core cooling? a.
RCS Thot is increasing uncontrollably.
i ' b.
RCS pressure is increasing uncontrollably.
c.
BOTH Steam Generator levels decrease to less than 70 inches.
f d.
RVLMS LVL 03 indicates voiding.
' '
! QUESTION: 058 (1.00) WHICH ONE (1) of the following is the Technical Specification basis for the-requirement to cooldown Unit 2 to below 500 degrees F within six (6) hours
when specific activity of the coolant exceeds 1.0 microcurie / gram DOSE l EQUIVALENT I-131 for greater than'48 hours?
! a.
Reduces the dissolution of fission products in the reactor. coolant.
b.
Increases reliability of the data collected for actual' iodine ( determination.
! c.
Reduces the rate of release if a steam generator tube rupture should l occur.
, d.
Increases coolant density sufficiently to enable self shielding thereby reducing on-site exposure.
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SENIOR REACTOR OPERATOR 'Page 39' , I i QUESTION: 059 (1.00) The following conditions exist: 1.
Unit 2 is at 30% power and increasing at 3% per hour.
2.
CEAs are in MANUAL SEQUEN'I'IAL.
3.
GROUP 5 CEAs were withdrawn from 50 to 60 inches.
4.
At 60 inches Group 5 CEAs continued stepping out.
i WHICH ONE (1) of the following actions are required to be performed FIRST per ! 2203.003, "CEA Malfunction"'r ! a.
Initiate emergency boration per 2203.032, " Emergency Boration".
b.
Attempt to stop CEA movement by placing MODE SELECTOR switch to the . OFF position.
! c.
TRIP the reactor and go to 2202.001, " Standard Post Trip Actions".
d.
De-Energize both MG sets by opening breakers 2B712 and 2B812.
>
QUESTION: 060 (1.00) , The following plant conditions exist: [
1.
Mode 3, Trip from 100% power occurred 5 minutes ago.
' i 2.
CEAs #40 and #41 are fully withdrawn.
3.
Charging pump suction piping has ruptured.
WHICH ONE (1) of the following operator actions should be performed under- ! these conditions? , a.
Use auxiliary spray and emergency borate.
b.
Emergency borate via HPSI by opening 2CVC-115.
j c.
Lower RCS temperature and commence a normal boration.
d.
Lower RCS pressure and emergency borate via HPSI.
! ,
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SENIOR REACTOR OPERATOR Page 40. QUESTION: 061 -(1.00) Considering Procedure 2202.009, " Functional Recovery Procedure", _ WHICH ONE - (1) ! of the following Safety Functions has the HIGHEST order of priority? a.
RCS Inventory b.
Vital Auxiliaries c.
Core Heat Removal d.
Containment Integrity i > QUESTION: 062 (1.00) , The following plant conditions exist: + ' 1.
Unit 2 has tripped from 100% power.
. 2.
SIAS has actuated.
3.
RCS pressure is 1300 psia and DECREASING slowly.
4.
Pressurizer level 1s 100%. ~ ' 5.
RCS Thot-(Th) is 557 degrees _F.
6.
Containment pressure is 16 psia and' INCREASING slowly.
WHICH ONE (1) of the following events has occurred?
a.
Main Steam'line break inside Containment b.
Inadvertent SIAS , c.
RCS cold leg break
d.
Pressurizer vapor space break f !
t , h . -.. _ _.- SENIOR REACTOR OPERATOR Page 41' , QUESTION: 063 (1.00) WHICH ONE'(1) of tne following'provides MAXIMUM cooling to the core for.the
first twelve-(12) hours during a SMALL BREAK'(300 gpm with RCS pressure
falling below 1250 psi) Loss Of Coolant Accident'(SBLOCA)? + _ -! a.
Reflux boiling l b.
Break flow' cooling c.
Natural Circulation d.
Forced Circulation ,
3 QUESTION: 064 (1.00) WHICH ONE (1) of the following parameters is used to determine if adequate _ core cooling exists following a small break loes of coolant accident-(SBLOCA)' per 2202.003; " Loss Of Coolant Accident, Safety Function Status Check"?. a.
RCS Thot and average CET temperatures are less than superheated.
! b.
Loop delta T is less than ten (10). degrees F.
, , c.
CET temperatures less than 1000 degrees'F.
d.
RCS Tcold is less than' ten (10) degrees F. superheated.
i a
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_ _ _ . . _ ~.. _ _ .. _ _ _ _ _.- .. ._ SEWIOR REACTOR OPERATOR Page 42-l QUESTION: 065 (1.00) WHICH ONE (1) of the following is the reason simultaneous Hot and. Cold Leg ' Injection is NOT established prior to two (2) hours following a' loss of l . coolant accident - (LOCA) ? a.
To avoid possible entrainment of SI flow in the Hot Leg due to steam 1 flow which is substantial prior to two. (2) hours.
b.
Boron precipitation is not a problem prior to two -(2) hours due'to-the large steam flow through the Hot-Leg out the break.
c.
Maximum Cold Leg flow is needed-to cool the core due to the high amount of decay heat _ prior to-two (2) hours.
d.
To allow time for vessel head cooling to prevent the possibility for _ void formation in the vessel head which would substantially reduce-core cooling.
QUESTION: 066 (1.00) The following conditions exist: 1.
A large break LOCA condition exists.
2.
RCS pressure is 140' psia.
' 3.
CETs indicate 267 degrees F.
4.
LPSI pump 2P-60A is injecting into the RCS.
5.
HPSI pumps 2P-89A and 2P-89B are injecting into the RCS.
- WHICH ONE-(1) of the following indicates the lowest acceptable total Emergency
, Core Cooling System flow per 2202.010, " Standard Attachments"? (References attached.)
a.
750 gpm.
' b.
1455 gpm c.
2125 gpm
d.
2900 gpm >
9 f y,, ,-rv , y,,- -r, - - + - - - - - - - - - - - - - - - - - - - - - - - - - - - -
..... ... - -. - - -..- - . _ - - -. . .. - - .. .. . SENIOR REACTOR OPERATOR Page 43 , .f QUESTION: 067 (1.00) .The following conditions exist after a Unit 2 reactor trip from 100% power: 1.
Pressurizer pressure is 450 psia and DECREASING.
2.
Pressurizer level is 0%. 3.
CETs indicate 370 degrees F.
4.
All ESFAS Actuations required have actuated.
WHICH ONE (1) of the following is the MINIMUM operator action for the above conditions per 2202.001, " Standard Post Trip Actions"? (References are provided.)
a.
Trip RCPs 2P32A and 2P32B.
b.
Trip RCPs 2P32C and 2P32D.
. c.
Trip RCPs 2P32A and 2P32C
d.
Trip ALL RCPs.
.
A QUESTION: 068 (1.00) WHICH ONE (1) of the following is the MINIMUM condition that defines' a " Harsh"
containment? a.
Containment temperature is 200 degrees F.
b.
Containment temperature is 150 degrees F. and pressure is greater than 8.5 psig.
c.
Containment humidity is saturated.
i d.
Containment humidity is saturated and Containment radiation ~ levels i indicate 1 X 10E3 R/hr.
- _.
' ..- .- .-. . .- - -.. ... . - .. - ,
, - .. . . - - ... . ~. . -, i i . SENIOR REACTOR OPERATOR Page144 ' ej QUESTION: 069 ( 1. 0 0 )'- ! WHICH ONE (1) of the following temperature indications-is preferred for , t monitoring RCS temperature following a loss of Shutdown-Cooling. flow? a.
LPSI pump discharge monitor (T5096) > ' b.
RVLMS ATS - c.
Hot Leg RTDs d.
CETs- ., . - . QUESTION: 070 (1.00) LThe'following conditions exist: 1.
Unit 2 is in REFUELING.
! 2.
Shutdown Cooling System is in service.
3.
One LPSI pump is operating.
. 4.
Unit 2 is drained down for nozzle dam. installation.
WHICH ONE (1) of the following Shutdown Cooling (SDC) parameters is an indication of a loss of adequate core heat removal per 2203.029, " Loss of-Shutdown Cooling"? ^ Shutdown cooling flow is 1500 gpm.
.. a.
I b.
Component Cooling Water (CCW) flow to the shutdown cooling 7 heat , exchangers is 1000 gpm.
c.
RCS Local Level indicator indicates 36 inches from the bottom of the- ~ ' hot leg.
! d.
RVLMS' indicates sensors 04 through 08 are WET.
' , , ' ' -c., .*w , - - = r
. -, -. ~ --.- .- ... .- - _. - - -. _ .- .._ , SENIOR REACTOR OPERATOR Page 45
, t QUESTION: 071 (1.00) WHICH ONE (1) of the following is the reason MG set supply breakers 2B712 and 2B812 are opened for a minimum of ten (10) seconds before re-closing when responding to an Anticipated Transient Without Scram (ATWS)? t a.
To assure all CEAs have sufficient time to fall into the core, b.
To allow sufficient time for the TCBs to open.
c.
To allow sufficient time for the MG sets to coast down and overcome the flywheel device.
d.
To prevent overcurrent trips when breakers are re-closed.
QUESTION: 072 (1.00) The following plant conditions exist: , 1.
Unit 2 is in REFUELING.
I 2.
Start-up channel "A" is out of service.
3.
Start-up channel "B" is OPERABLE.
i 4.
Core alterations are in progress.
i ! WHICH ONE (1) of the following MINIMUM Technical Specification Action ] ' . Statements should be implemented? i a.
Suspend all operations involving positive reactivity changes, b.
Determine the RCS boron concentration at least every twelve (12) hours.
c.
Immediately evacuate containment until the audible alarm from Channel "A" is returned to service.
d.
Perform CHANNEL FUNCTIONAL TEST on Channel "B" within one (1) hour.
j .- - - -,, - - -- - - . -. . - -.
_.
__ _ _.
.... _ _ -.- _ _ _ _ . _.... _. _. __ __ _.. _ ' SENIOR REACTOR OPERATOR Page 46 , ':
- QUESTION: 073 (1.00)
The following conditions exist: 1.
Unit 2 has reduced power to 15% due to a Steam Generator tube leak.
2.
Letdown has been isolated.
3.
One (1) charging pump is running.
4.
Pressurizer level is DECREASING slowly.
~ WHICH ONE (1) of the following is a valid method to determine the amount of primary to secondary leakage? a.
Use Charging flow and Letdown flow mismatch, b.
Use SG Tube Leak N-16 monitors.
c.
Use Charging flow minus Controlled Bleedoff flow.
, d.
Use Main Steam Line' Radiation monitors.
QUESTION: 074 (1.00) Unit 2 is being shutdown to HOT STANDBY due to a steam generator tube leak.
WHICH ONE (1) of the following indicates the MINIMUM conditions necessary for-a manual reactor trip per 2203.038, " Primary to Secondary Leakage."? a.
When all available charging pumps are running, letdown has been i isolated, and pressurizer level can no longer be controlled at setpoint.
! b.
When all available charging pumps are running and letdown has-i reached minimum flow.
..) c.
When letdown flow combined with the tube rupture flow exceeds 132 ' GPM with three (3) charging pumps running.
d.
When charging flow reaches 132 GPM.
j , ...-. . . .... .. . _ -.. - . -.. -, -- . .. - .. _. - - _ _, -
. _ _ _ .., _ _. _ . - - _ - _ _ _ _ _ .. _.. SENIOR REACTOR. OPERATOR Page 47 . QUESTION: 075 (1.00) The following conditions exist: ~ 1.
Unit 2 has tripped from 100% power due to a Steam Generator tube' , rupture on the "B" Steam Generator.
2.
Operators have throttled HPSI flow to. assist in' recovery operations.. 3.
RCS pressure is 1800 psig.
WHICH ONE.(1) of the following conditions will require operators tx) restore FULL HPSI flow? -a.
Margin to saturation is 40 degrees F.
b.
Pressurizer level is 25%. c.
RVLMS sensor four (LVL 03) indicates WET.
d.
Steam Generator "A" level is 5% and increasing with EFW flow of 500 gpm.
-
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.~ _ . -. .. . . -._.
-_._.
. . .. _ _ _ SENIOR REACTOR OPERATOR Page 48 QUESTION: 076 (1.00) The following conditions exist: 1.
Unit 2 has been MANUALLY tripped following a rupture of Steam Generator "B".
2.
SIAS has automatically actuated.
3.
All RCPs are secured.
- 4.
RCS pressure is 1000. psia.
5.
CETs indicate 450 degrees F.
6.
Pressurizer level indicates 26%. 7.
RVLMS LVL 01 indicates WET.
WHICH ONE (1) of the following actions must be performed prior to resuming RCP 2P32C operation per 2202.004, Steam Generator Tube Rupture"? " a.
Seal injection must be supplied to RCP 2P32C.
t b.
RCP 2P32B must be started first to establish Pressurizer Spray flow.
c.
RCS subcooling must be increased to within allowable RCP restart' limits, d.
Pressurizer level must be increased to within allowable RCP restart limits.
l l l , y ,,, - - .-,. u 1.*e ,, ,- ~. - -. - -- - -, --.. .-- . - - - - - - - --
. _ _ _ _____ _ _.
__ . _.. _.. - _ _ _. __ ._ _ _ ._._ _ _ _ _ SENIOR REACTOR OPERATOR Page 49 QUESTION: 077 (1.00) The following conditions exist: 1.
Unit 2 has tripped from 100% power when ALL Feedwater is lost.
2.
The Turbine Driven Emergency Feedwater pump 2P7A can be returned to ' service.
3.
Steam Generator "A" level indicates 20%. l 4.
Steam Generator "B" level indicates 46%. WHICH ONE (1) of the following is the correct method for re-establishing feedwater flow to the Steam Generators? a.
Initiate feedwater to the "A" Steam Generator at the maximum flow rate possible for the first five (5) minutes.
b.
Initiate feedwater to'BOTH "A" and "B" Steam Generators at the naximum flow rate possible for the first five (5) minutes.
, c.
Initiate feedwater slowly to "B" Steam Generator at a maximum flow rate of 150 gpm for the first five (5) minutes.
, d.
Initiate feedwater slowly to BOTH "A" and "B" Steam Generators at a maximum flow rate of 150 gpm for the first five (5) minutes.
l ' QUESTION: 078 (1.00)
The following conditions exist: 1.
Unit 2 has been manually tripped due to a loss of ALL Feedwater.
2.
Operators are initiating step 3 of 2202.006, " Loss of Feedwater" and are tripping ALL RCPs.
S WHICH ONE (1) of the following is the basis for tripping all Reactor Coolant-Pumps as per this step in the procedure? a.
To reduce heat input into the reactor coolant system.
b.
To reduce Reactor Coolant System pressure.
I c.
To reduce thermal stress to steam generators.
d.
To reduce-tube to shell delta pressure.in the steam generators.-- !
.~.-.... m .,.. - , ,_ g , -+, e-
. -.. -.. - _. . _ _ _ _ _ _ . _ _ _ _ _ _ _.. . _ _ - SENIOR REACTOR OPERATOR Page:50 '1 QUESTION: 079 (1.00) LWHICH ONE (1) of the following is the MAXIMUM time that the Vital 125 VDC-batteries 2D-11 and 2D-12 can be expected to carry.the vital loads following a.
~1oss of the battery chargers? a.
2 hours b.
8 hours c.
16 hours , i d.
24 hours P QUESTION: 080 (1.00) The following conditions exist: 1.
Unit 2 is at 100% power.
2.
Pressurizer level control is selected to Channel "A".
3.
Proportional Heater Bank handswitch is in AUTOMATIC.
j l WHICH ONE (1) of the following is the affect of a loss of power supply 2Y1 on-the' Pressurizer Level Control System? a.
Channel "B" instrumente FAIL.
b.
Proportional and Backup Pressurizer heaters are ENERGIZED.
. c.
Proportional and Backup-Pressurizer heaters are DE-ENERGIZED.
d.
Pressurizer level indication at the Remote Shutdown Panel is' LOST.
' , .v,. ... . _, _ _ -_ ._,.,.. . ._, , , _ _.., _ _, .
. _-. _ _ _ . -._. _.. _ - i I SENIOR REACTOR-OPERATOR Page 51 QUESTION: 081 (1.00) ! The following plant conditions exist: 1.
Core alterations are in progress on Unit 2.
2.
A spent fuel assembly is in the hoist box on the Main Refueling bridge, withdrawn 30 inches from the core.
3.
A Station Blackout then occurs.
WHICH ONE (1) of the following actions should be performed? a.
Leave the fuel assembly and evacuate containment immediately.
-, l b.
Hand jack the fuel assembly back into-the core, j c.
Hand jack the fuel assembly into a storage rack.
d.
Hand jack the fuel assembly completely in[o the hoist box,.then hand: jack to the spent fuel pool.
l QUESTION: 082 (1.00) WHICH ONE (1) of the following is the MINIMUM level required in the Steam Generators in order to sustain Natural Circulation Cooling (NC) ? , a.
125 inches b.
200 inches c.
275 inches d.
350 inches ~ ; . ; ,
F > -- -,-,c, e e-, , , ., - .---e a-n .', ~N . -. - - -. - - - - - - -.. - - n-
... _. .m.. . . _ ..-_ _... - _ , _.. _.-. . _. _ . _ -. _ - . -SENIOR REACTOR OPERATOR' Page.52 l QUESTION: 083 (1.00) WHICH ONE_(1) of the following valves will fail CLOSED on a loss'of Instrument Air pressure? a.
2CV-0634,-Hotwell Makeup valve b.
2CV-1051, Steam Generator "B" Upstream Dump valve c, 2CV-0740, FW Loop "B" Main Reg valve d.
2CV-5091, SDC HX Flow Control valve , i- " QUESTION: 084 (1.00) WHICH ONE (1) of the following jobs can be performed under an Operations Standing Radiation Work Permit (RWP)? a.
An entry into a High Rad.iation Area where accumulated dose is expected to be 75 mrem.
b.
An entry into a High Contamination Area where accumulated dose is expected to be 10 mrem.
c.
A job which requires opening-a potentially contaminated steam line.
d.
A job which requires use of respiratory protection'due to loose'
contamination.
't ,
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.. , .- , -. - _ -..-.. -- -. _. ... --. --- - ., ... -.. - ~ SENIORfREACTOR OPERATOR 'Page 53 .
- QUESTION: 085 (1.00)
WHICH ONE-(1) of the'following atmospheric conditions defines an acutely disabling atmosphere? a.
Oxygen concentration is 20%. b.
Ammonia concentration is 5 PPM.
c.
Chlorine concentration is 5-PPM.
d.
Hydrazine concentration is.05 PPM.
re QUESTION: 086 (1.00) The following conditions exist: 1.
A planned job is being conducted in a Confined Space.
i 2.
Several crews are working the job on a rotating basis approximately every two (2) hours.
, " 3.
The job is to last around the clock for two (2) days.
, WHICH ONE (1) of the following is the MINIMUM frequency required for issuing a' , Confined Space Entry Permit? a.
The permit is issued once by Planning and Scheduling prior.to.the-
start of the job.
' b.
The permit is issued once when the first crew enters the -job site.
. c.
A new permit is issued every 6 hours if ventilation'is maintained, d.
A new permit is issued-each time a crew enters the job' site, a , . P , L n-n, r,. w, ~ + -.
.. _ _ .. _ _.. _ _ . . _ _ _ _.
_ _ _ _. - _ _ _. ~ -.. _.. ~ _ _,. i . SENIOR. REACTOR OPERATOR Page 54 , QUESTION: 087 (1.00) WHICH ONE (1) of the following is designated to perform an emergency rescue of.
' an individual in a Confined Space?
a.
Any individual who has received training as an ATTENDANT.
b.
Any individual who has received training to make confined' space-entries.
c.
Any member of the Bechtel Emergency Team.
d.
Any member of the Pope County Rescue Team.
, QUESTION: 088 (1.00) The following conditions exist: 1.
Unit 2 is operating at 100% power.
2.
The latest Chemistry sample indicates that the RCS has a dissolved Oxygen concentration of 1500 ppb.
WHICH ONE (1) of the following MINIMUM actions is applicable for these conditions? a.
The Action Level 1 limit has been exceeded which requires Oxygen.
concentration to be reduced to less than five (5) ppb within seven (7) days, b.
The Action Level 2 limit has been exceeded which. requires Oxygen concentration to be reduced to less than five (5) ppb within twenty four (24) hours.
The Action Level 3 limit has been exceeded which requires' Oxygen c.
concentration to be reduced to less than one hundred (100) ppb.
within twenty four (24) hours, d.
The Action Level 3 limit has been exceeded which requires the.
reactor be shutdown immediately and RCS temperature reduced to less-than 250 degrees F.
as soon as possible.
. ,.,...., _ _.. _ _ ~.. - '.. - -. .-;, . -,,.. ~ -,, , ... . ,, _.. _ _ .
- . -. - - . ~.. - ...- ...- -. - . ... -. -.. SEMIOR REACTOR OPERATOR Page 55 ~ ' QUESTION: 089 (1.00) WHICH ONE (1)'of the following boundaries is challenged the mostLby a chloride-excursion in the RCS? a.
Fuel cladding b.
RCS letdown piping , c.
RCS cold leg piping d.
Containment liner -
. QUESTION: 090 (1.00) WHICH ONE (1) of the following is an indication that an electrical _ hand tool issued from the toolroom on 12-13-1993 was recently inspected and is safe for-use? a.
ANY piece of equipment issued from the toolroom is assumed to be safe for use.
b.
The tool will be' tagged with a orange tag, c.
The tool's cord will be marked-with a continuous ring of orange electrical tape behind the plug.
d.
The tool will have an orange inspection sticker affixed to it's , surface.
' ,
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_ . .. - . . --.., - . _... _ _ _.... -~ - 1, SENIOR-REACTOR OPERATOR' Page 56 . .. ; QUESTION: 091 (1.00) 'WHICH ONE (1) of the following designates a CONTINUOUS ACTION STEP in'an Emergency Procedure? a.
An asterisk precedes the step.
. ' ' b.
An solid square precedes the step.
c.
Step number is in brackets.
' d.
Step number is BOLD.
, QUESTION: 092 (1.00) - WHICH ONE (1) of the following is the MAXIMUM time-allowed'to report an - UPGRADE in an event classification to the NRC? a.
five (5) minutes b.
fifteen (15) minutes iq c.
thirty (30) minutes t d.
sixty (60) minutes-I Y ' k
9
- i
' ..,. -. ...,,. .... -...a.
._ . ... -. _. _. _.
..,
- - . _- - _.
.. _.. . -. SENIOR REACTOR OPERATOR Page 57 QUESTION: 093 (1.00) The following conditions exist: 1.
Hold Cards are being removed from equipment in the Auxiliary " Building.
2.
A Licensed Control Room Operator is in contact with the crew in the Auxiliary building by telephone.
WHICH ONE (1) of the following is the MINIMUM amount of documentation needed on the Hold Card Record Sheet if the tag removal is being documented PER-TELECON? The Licensed Operator needs only to initial the sheet with his own a.
initials and enter Telecon into the Remarks Section of the Hold Card Record Sheet, b.
The Licensed Operator needs only to initial the sheet with the initials of the person removing the tags.
c.
The Licensed Operator needs to initial the sheet with his own initials and those of the person removing the tags and enter Telecon into the Remarks Section of the Hold Card Record Sheet.
d.
The Licensed Operator needs to initial the sheet with his own initials and the Shift Supervisor must initial those of the person removing the tags.
m -_ _ _ . _. _. _. _. _ _ _, _.. -. _. _.. _. _ _.. .a
_ , . _. _ _. _.. _ _ _ _ _.
.... . ._.. _ _ _ -. _. _ SENIOR REACTOR OPERATOR Page 58 , .; , QUESTION: 094 (1.00) The following conditions exist: , 1.
A crew is preparing to conduct MOV testing.
2.
They find a HOLD CARD tagging the breaker OPEN for the valve they are to test.
WHICH ONE (1) of the following actions should be taken next? a.
Install the MOV TEST CARD over the HOLD CARD and proceed with MOV testing.
, b.
Proceed with MOV testing under protection of the HOLD CARD and a-MOV TEST card is not needed.
c.
Install the MOV TEST CARD but contact the Shift Supervisor for-further instructions.
d.
Contact the Shift Supervisor for authorization to remove the HOLD CARD prior to installing the MOV TEST CARD.
QUESTION: 095 (1.00) WHICH ONE (1) of the following represents an INTENT Change in a procedure? a.
Deleting a step that closes a valve that has been removed by a Design Change.
b.
Correction to a typographical error in an ACA.
, c.
Deleting a step that requires a hold-point.
' d.
Modification to step sequence.
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. _. _ . _.... - _. _. _..._. _ . . .. -.. - _ . _.. _. _ -.. A SENIOR REACTOR' OPERATOR Page 59
QUESTION: 096 (1.00) The following crew members are present-to assume control of the shift while the Unit is operating at 100% power: , 1.
One (1) SRO 2.
Three (3) ROs 3.
One (1) Waste Control Operator 4.
One (1) AO ' 5.
One (1) STA WHICH ONE (1) of the following additional crew members is needed to complete'a.
MINIMUM Shift Complement per 1015.001, " Conduct of_ Operations"?
a.
One (1) SRO b.
One (1) RO . i c.
One (1) Waste Control Operator d.
One (1) AO , QUESTION: 097 (1.00) WHICH ONE (1) of the following is the MINIMUM level of approval'needed to "N/A" a procedure step in an approved operating procedure? a.
Two (2 ) Licensed Reactor Operators b.
Two (2) Licensed. Senior Reactor Operators c.
Shift Engineer AND Control Room Supervisor.
d.
Shift Superintendent , I , v < ., - . - - - -., - -- ,,. . . .n,
.. . . . - - - . -- . -... ..... , SENIOR' REACTOR. OPERATOR Page 260 QUESTION: 098 (1.00) WHICH of the following individuals may be issued a key to the Containment Personnel Hatch? a.
A qualified Reactor Operator b.
A Security Officer c.
A qualified Senior Reactor Operator d.
The Health Physics Supervisor
QUESTION: 099 (1.00) The following conditions exist: 1.
Unit 2 is performing Refueling Operations.
2.
The Shift Supervisor needs a control room operator for the upcoming evening shift on a Saturday night to replace an 111~ crew' member.
. WHICH ONE (1) of the following is the PREFERRED operator to fill this vacancy , per 1015.001, " Conduct Of Operations"? ! a.
Operator "A", who has worked his normal dayshift, is willing to hold over on to the evening shift but he states that he came in one hour-early to relieve an operator for a doctors appointment.
b.
Operator "B", who has' worked his normal dayshift (8 hours), is willing to hold over on to the evening shift but he states that he . worked eight (8) hours on the evening shift the day before after , working day shift.
' c.
Operator "C", who works relief dayshift, is willing to work the - evening shift but states that he worked an overtime shift.two (2) ' days ago and it would be his third overtime shift this weeki d.
Operator "D", who works the midnight shift, is willing to cnme'in and. work the evening shift, but states he held over from his previous shift for a'one (1) hour training class.
, ,, - - r -.
.
' SENIOR REACTOR ONERATOR Page 61 ' -QUESTION: 100 (l'. 0 0 ) . WHICH ONE (1) of the following will meet the MINIMUM number of Fire Brigade Trained personnel that are required to respond to a fire.on Unit 2 per Administrative Procedure 1015.007, " Fire Brigade Organization and-Responsibilities", (including Brigade Leader) ? , a.
Two (2) operations-personnel from Unit 2 and one (1) security representative b.
Two (2) operations personnel from Unit 2, one (1) security representative, and one (1) operations representative from Unit 1 c.
Two (2) operations personnel from Unit 2, Two (2) security-representatives, and one (1) operation's representative from Unit 1 d.
Two (2) operations personnel from Unit 2, Two (2) security.
representatives, and two (2) operation's representatives-frbm Unit 1 l , l i l l .] ! (*******,, END-OF EXAMINATION ******,,,,) l - _. .. - - - - - -- - -. ... .- -. .
, ._ __.m .__ ._.. -... . . - _. . _ _. - ... _ _.. _..... _ -. _.. . _ _...- & - SENIOR REACTOR OPERATOR Page 62 . . ANSWER: 001 (1.00) , c.
(+1.0]
REFERENCE: + 1.
2203.032, " Emergency Boration", page 3.
. 2.
AA52102-003, " Chemical & Volume Control System", Objective 8.
i 3.
KA 004000K205 (2.7/2.9) 004010A207 (3.8/3.9) 004010A403 (3.9/3.7) 004010A403 004010A207 004000K205 ..(KA's)
, ANSWER: 002 (1.00) b'. [+1.0) - REFERENCE: 1.
AA52002-014, "Excore Nuclear Instrumentation", Objective 14.3f, page 29, 2.
KA 015000K106 (3.1/3.4) 015000K106 ..(KA's) i ANSWER: 003 (1.00) l , b.
[+1.0) = REFERENCE: ' 1.
AA-52002-012, " Control Element Drive' Control", Objective 12,6_,-page 16.
2.
KA 001000K105 (4.5/4.4) 001000K105 ..(KA's) '
- ANSWER: 004 (1.00) - .a.
[ + 1. 0 ) - , ,
-l i , ,, .. . . - - - ... , . . -..
..... -., . .. - - . _ - ~ . -. -. ~ . .. . ~. -.... -. - .-. sSENIOR REACTOR' OPERATOR Page 63 - -REFERENCE: 1.
'AA52002-001, " Reactor Coolant System", Objective 1.6, page.41.
2.
KA 003000A402 (2.9/2.9) ' -) 003000A402 ..(KA's) . ANSWER: 005 (1.00) , d.
(+1.0) i-REFERENCE: t 1.
AA52002-001, " Reactor Coolant System", Objective 1.7, page 41, 2.
KA 003000G013 ' (3. 6/3. 7), 003000G010 (3.3/3.6) 003000G010 003000G013 ..(KA's) l J r ' ANSWER: 006 (1.00) d' (+1.0) . i REFERENCE: l 1.
.AA-52002-013, " Engineered Safety Features Actuation System", Objective 13.2B, page 16.
2.
KA 013000A301 (3.7/3.9) 013000A301 ..(KA's) . . ANSWER: 007 (1.00) P c.
[+1.0] REFERENCE: g 1.- AA520021037, " Saturation Margin Monitor", Objective 37.8, page 9.
2.
KA 017000G010 (2.6/2.9) .i 017000G010 ..(KA's) r
, .. .- - .... . .... .- .,. - _ - .. _
SENIOR 2 REACTOR OPERATOR Page 64 i ANSWER: 008 (1. 0 0 ) ' c.
[+1.0) .. REFERENCE: 1.
STM-2-09, " Containment' Cooling and Purge System", page 1.
ii 2.
AA52002-027, " Service Water System", Objective 27.2J.
. 3.
.KA 022000K101 (3.5/3.7) - -022000K101 ..(KA's) , ANSWER: 009 (1.00) ' d.
[+1.0) ' REFERENCE: i
1.
AA32003-021, " Condensate and Feedwater System", Objective 21.6,.page 41.
2.
KA 059000G010 (2. 9/2. 9 )
059000G010 ..(KA's) f ANSWER: 010 (1. 0 0) ' c.
[+1.0) , ' REFERENCE: 1.
AA32003-021, " Condensate and Feedwater. System", Objective 21.3h,.page 31.'. .2.- KA 059000K419 (3.2/3.4) _ .! t 059000K419 ..(KA's) . <' _, . . - . _, - .. - --. . .... z z '
, . . -.. - ~... ,. -.-. - -. -. - - .. -.. .. ...... - _. -... -..... i . . I . SENIOR REACTOR OPERATOR Page 65 ' ANSWER: 011' (1.00)- -l b.
[+1.0) i REFERENCE: ' 1.
AA 42002-021, " Emergency'Feedwater System", Objective ~21.2B,-page 6.
2.
KA 061000K401 (3.9/4.2) 061000K401 ..(KA's) i ' ANSWER: 012 (1.00) c.
[+1.0) - REFERENCE: ] -i '1.
AA-520022013, " Engineered Safety Features Actuation System", Objective , 13.2D, page 20.
2.
KA 061000K414 (3.5/3.7) 061000K414 ..(KA's) , . ANSWER: 013 (1.00) c.
[+1.0) REFERENCE: 1.
AA42002-021, " Emergency Feedwater System", Objective 21.3C, page 14.
2.: 101 061000A303 (3.9/3.9) ' 061000A303 ..(KA's) ' . ANSWER: 014 (1.00) d.
[+1.0) , ,. - - _. -.. -_. - -. - . - ,. --
.... _.. _ m._.. _.. . _ _.. - i - SENIOR REACTOR OPERATOR - Page 66.
' , REFERENCE: ' 1.
AA-52002-033, "Radwaste System", Objective 33.4, page 27.
! 2.
KA 068000A302 (3.6/3.6) 068000A302 ..(KA's) , - ! ANSWER: 015 (1.00) i a.
[+1.0) , REFERENCE: ' . i ~1.
STM-2-52, " Liquid Waste System", page 5.
2.
AA-52002-033, "Radwaste System", Objective 33.4.
3.
KA 068000K107 (2.7/2.9) 068000K107 ..(KA's)
i
ANSWER: 016 (1.00)
c.
[+1.0) REFERENCE: - 1.
2104.010, " Unit 2 Waste Gas Analyzer. Operation", Attachment "A",.page 23.
2.
AA-52002-033, "Radwaste System", Objective 33.4, page 32.
2.
KA 071000K504 (2.5/3.1), 071000A429 (3.0/3.6) i , , 071000A429 071000K504 (KA's) .. ANSWER: 017-(1.00) d.
[+1.0} ,
s
REFERENCE: 1.
AA-52002-018, " Radiation' Monitoring System", Obj ective: 18.5, page 9.
2.
KA 072000G010 (2.8/3.0)
072000G010- ..(KA's) R-e , - +.. ~.. .- , - - ,c .. +.v., ,
.. _ _. ._ .. . _ _ _.. _ _.
. . _. _ .. _ ' SEMIOR REACTOR OPERATOR Page 67 i ANSWER: 018 (1.00) a.
[+1.0) REFERENCE: 1.
AA52002-001, " Reactor Coolant System",. Objective 1.2, page 10.
2.
KA 002000K108 (4.5/4.6) 002000K108 ..(KA's) ANSWER: 019 (1.00) . ' b.
[+1.01 s - REFERENCE:
1.
Technical Specifications 3.4.6.2, " Reactor Coolant System' Leakage", page 3/4 4-14.
2.
AA52002-001, " Reactor Coolant System", Obj ective 1.16.
3.
KA 002000G011 (3.3/4.0) NOTE: Answer 10. 5 - ( 2. 5 +2. 2 +2. 5 + 1. 2 ) - 2.1 gpm unidentified which is.> . ' 1.0 gpm unidentified limit.
! 002000G011 ..(KA's) ' , , ' ANSWER: 020 (1.00) a.
[+1.0) REFERENCE: , 1.
AA52002-004, " Emergency Core System", Objective 4.9A, page-17.
2.
KA 006020A107 ( 3 ~. 5 /3. 7 ) , ! 006020A107 ..(KA's) f
1 r , .. - , ,. -., _. -.,.. , ..n , . . -.. --,,,,
.... ,.... _... _. -._. . - ....-.~.......___.___.s.
......., ' SENIOR-REACTOR.OPERATO'R Page 60' ' ANSWER: -021 (1.00) c.
[+1.0] I
REFERENCE: , l'. -AA52002-004, " Emergency Core System", Objective 4.4,.page'9.
'2.. KA 006000K603 (3.6/3.9) 006000K603 ..(KA's) i , < ANSWER: 022 (1.00) a.
[ + 1.' 0 ] . . REFERENCE: , 1.
2103.005, " Pressurizer Operat. don", page 9.
-2.
AA52102-002, ~STG2A, Objective 1.1, Attachment A, page 32 of 45, 3.
AA52002-001, " Reactor Coolant System", Objective 1.11, page 60.
4.
KA 010000A402 (3. 6/3.4), 010000A101 (2. 8/2. 9 ) + 010000A101 010000A402 ..(KA's) , ! ANSWER: 023 (1. 0 0 )' a.
.[+1.0) , . REFERENCE: 1.
AA52002-001, " Reactor Coolant System", Objective 1.9, page 50.
2.
KA 011000K605 -l(3.1/3.7) 011000K605 ..(KA's) , ANSWER: 024 (1.00) , d.
[+1.0) ,
e n , , , w w- ,w.
.,, -,. -m, .e -, , e n .-e>,
<
, i
SENIOR REACTOR OPERATOR.
.Page 69 REFERENCE: 1.. AA-52002-043, " Diverse Scram System", Objective 43.10, page 8.
2.- KA 012000K201 (3.3/3.7), 012000K107 (3.2/3.2) . I ) ' 012000K103 (3.7/3.8) 012000A407 (3.9/3.9) 012000A407 012000K103 012000K107 012000K201 ..(KA's).
ANSWER: 025 (1.00) c.- [+1.0] , ! REFERENCE: 1.
AA52002-037, " Saturation Margin Monitor", Objective 37.6, page 8, 2.
KA 016000A302 (2.9/2.9) . ,# , , 016000A302 ..(KA's) ANSWER: 026 (1.00) c '. [+1.0] ! REFERENCE: . , 1.
AA42002-007, " Containment Spray System", Objective 7.4.
2.
STM 2-08, " Containment Spray System", page 3.
3.
KA 026000K201 (3.4/3.6) .; 026000K201 ..(KA's) ANSWER: 027 (1.00) . . d.
I+1.0) , i . + , B
- - . . . -
..._ ., - . ... - - .-. < ... f .- ' SENIOR REACTOR OPERATOR Page 70 , LREFERENCE: , '1.. AA-52002-013, " Engineered Safety Features Actuation System", Objective ' '13.2D, page.15.
. . ! 2.
KA 026000A301 (4. 3/4. 5), 026000K101 ( 4. 2 /4.'2 ) 026000A301-026000K101 ..(KA's) . t " ANSWER: 028 (1.00) , c.
[+1.0) REFERENCE: . 1.
STM-2-09, " Containment Cooling and Purge. System", page 7.
' 2.
KA 029000K104 ( 3. 0 / 3.1) 029000K104 ..(KA's) - ANSWER: 029 (1.00) , .b.. {+1.0) REFERENCE: 1.
2-07, "Spes '- Fuel System", page 1.
2.' - .2001-008, L! 'tive 8.1.
3.
Ka 033000G010 2/2.5) , 033000G010 ..(KA's) , ANSWERi 030 (1.00) i a.
[+1.0)
i . h . , - -. - .. ...
-. .. - ..., ^ .., l l SENIOR REACTOR OPERATOR Page 71 , 1 REFERENCE: -1 i 11.
AA52102-015, "Feedwater' Control System", Objective 4, page.18.
. J l2.
KA 035010A301 (4.0/3.9) 035010A301 ..(KA's) ., ' ANSWER: 031 (1.00) a.
[+1.0) . REFERENCE: 1.
AA52102-015, "Feedwater Control System", Objective 7, page 27.
' 2.
KA 035010A203 (3.4/3.6) -1 035010A203 ..(KA's) >
! ' ANSWER: 032 (1.00) d.
[41.0) i ' REFERENCE: 1.
STM-2-15, " Main; Steam", page 4.
. + 2.- 2106.016, " Condensate and Feedwater Operation", Supplement 1,-page 116, 3.
KA 039000K408-(3.3/3.4) j a 039000K408 (KA's) . , t ANSWER: 033 (1.00) . i d.
[+1.0) i .1 'f P ! 1-4 i . -.. - - - . -
... -.... .. -... - . - -.. ... ~ ~.,, -.. -.- -.-. ..-.-.....-n . - + . ' SENIOR REACTOR OPERATOR-Page 72 l , i < _. REFERENCE: 1~. AA52002-032, " Control Room and ESF Equipment Ventilation System", > Objective 32.2, page 5.
2.
.KA 073000G007 (2.9/3.0) 073000G007 ..(KA's) .;
ANSWER: 034 (1.00) b.
[+1.0] . REFERENCE: i 1.
AA52002-044, " Shutdown Cooling System", Objective 44.2', page 7.
'; 2.
KA 005000K110 (3.2/3.4) { 005000K110 ..(KA's)
,
ANSWER: 035 (1.00) d.
[+1.01 REFERENCE: 1.
AA52002-001, " Reactor Coolant System", Objective 1.8, page 62'. 2.
KA 007000A301 (2.7/2.9) 007000A301 ..(KA's)
ANSWER: 036 (1.00) c.
[+1.0)
? . .. . I i . i . . , t ,
- -_ _ ... .,. , .. ._._ _.
- s . SENIOR: REACTOR OPERATOR Page 73- 'l
REFERENCE: 1.
AA42002-007, " Containment Spray System", Objective 7.6, page 20.
2.
KA 027000G007 (3.2/3.4) 027000G007 ..(KA's)
ANSWER: 037 (1.00) b.
[+1.0) ' REFERENCE:
1.
AA52002-050, " Containment Combustible Gas Control", Objective 50.7,-page , 21.
2 '. KA 020000G010 (3.0/3.2) ' ~028000G010 ..(KA's) ' - ANSWER: '038 (1.00)- d.
[+1.0) ' - REFERENCE: ' 1.
AA-52002-006, " Fuel Handling Equipment", Objective 26.3, Attachment A, page 1.
2.
KA 034000K402 (2.5/3.3) -; ~ 034000K402 ..(KA's) l ANSWER: 039 (1.00) b.
[+1.0] , f
'- , - - . . . .
_ _ _ _ _ ! SENIOR REACTOR' OPERATOR.
Page 74- - , ' REFERENCE: 1.
AA-52002-027, " Service Water", Objective 27.6, page 29.
' 2 ~. - .KA 076000G005 (2. 8/3.2), 076000G010 (2.7/2.9) 076000G010-076000G005 ..(KA's) . ANSVFR.
040 (1.00) d.- (+1.0) , , REFERENCE:
1.
'22r3.021, " Loss of Instrument Air", page 2.
'2.. STM-2-48, " Instrument Air System", page 3.
3 '. KA 078000K303 (3.0/3.4) ! 078000K303 ..(KA's) ANSWER:- 041 (1.00) d.
[+1.0) _ REFERENCE: , , 1.
Technical Specification Bases-3/4.1.3, " Movable Control Assemblies", page-B 3/4 1-5.
2.
KA 000005K302'(3.6/4.2) 000005K302 ..(KA's) -ANSWER: 042 (1.00) a.. (+1.0] '
- REFERENCE
1.
2203.025, "RCP Emergencies", Attachment D, page 14.
, 2.
KA.000015G010 (3.4/3.4) ' 000015G010 ..(KA's) , , , m . . .. . -
_ - SENIOR. REACTOR' OPERATOR' Page 75-ANSWER: 043' '(1. 0 0 ) .c.
[+1.0} ' REFERENCE: - 1, 2203.025,'"RCP Emergencies", page 5.
2.
~KA 000015A122 (4. 0/4.2 ). 000015A122 ..(KA's) ANSWER: 044 (1.00) a.
[+1.0} ... ,
REFERENCE: '1 AA52002-014, " Reactor Trip Recovery", Objective 14.5, page 6 2.
KA 000024K302 (4.2/4.4) ' '000024K302 ..(KA's) ANSWER: 045 (1.00) 'd.
[+1.0] REFERENCE: 1.
Technical' Specifications Bases 3.5.4,." Refueling Water Storage", page'B 3/4'5-2, 2.
KA 000024K104 (2.8/3.6), 000024G004 :(2.9/3.8).
000024G004 000024K104 ..(KA's) e L ANSWER: 046. (1.00) c.
[+1.0) o
_ _ _ ., _ -_.
L.
+i SENIOR-:REA:iOR_ OPERATOR 'Page 76 ' '
- REFERENCE:
-1.. 2104.028, " Component Cooling Water System Operation", page 6.
Objective 3'.2H,.pages 15-17.
O 72.
AA52002-030, " Cooling Water System", 3.
KA 000026A201 (2.9/3.5) 000026K301 (3.2/3.5), 000026A102 (3.2/3.3) ' 000026A102 000026K301 000026A201 .. - ( KA ' s ) ,
ANSWER: 047 (1.00) , b.
[+1.0) .; REFERENCE: . -1.
AA42102-002, " Pressurizer Pressure and Level Control System", Q10.,.page 6.
2.
KA 000027K303 (3.7/4.1)y - i 000027K303 ..(KA's) ' ANSWER: 048 (1.00)
a.
[+1.0) n REFERENCE: -j 1.
AAS2003-009, " Excess Steam Demand", page 6.
'2.
KA 000040K106 (3.7/3.8).
000040K106 ..(KA's) ANSWER: 049 (1.00) d.
[+1.0) i REFERENCE: 1. : 2202.010, " Standard Attachments", Attachment 20, page 60.
2.
KA 000040A203 (4.6/4,7) 000040A203 ..(KA's) -.- .
. _, . _ _ -. _.. _._.m- - - - _ .. SENIOR REACTOR OPERATOR Page 77 . cANSWER: 050.
(1.00) b.
( + 1. 0) REFERENCE: 1.
2203.019, " Loss Of Condenser Vacuum", page 4.
2.
.KA 000051A202 (3.9/4.1) 000051A202 ..(KA's) . ANSWER: 051 (1.00) c.
[+1.0) e,,-- REFERENCE: 1.
AA52003-011, " Loss of Offsite Power", Objective ~11.6, pages 20-21.
2.- KA 000055A202 (4.4/4.6) 000055A202 ..(KA's).
ANSWER: 052 (1.00) a.
[+1.0) REFERENCE: 1.
AA52003-012, " Station Blackout", Objective 12.5, page 14.
2.
KA 000055G007 (3.6/3.7) .000055G007 ..(KA's) > ANSWER: 053 (1.00) v ' d.
{+1.0) . . . - . ... -
. . ... -_ .. .. - - _.. - ~ . -.._._ _ .. , SENIOR REACTOR OPERATOR Pagen78 UREFERENCE: 1.
2203.034, " Fire Or Explosion", page.2.
2.
KA 000067K102 (3.1/3.9) 000067K102 ..(KA's) i.i . ANSWER: 054 (1.00) b.
[+1.0] REFERENCE: 1.
T/S 3.3.3.5, Table 3.3-9, p. 3/4 3-37.
2.
KA '000068K201 (3.9/4.0) 000068K201 ..(KA's)- , ANSWER: 055 (1.00) 'b.
{+1.0) . ? REFERENCE: ' 1.
2203.014, " Alternate Shutdown", page 3.
2.
KA 000068G010 ( 4.' 1/ 4. 2 ) 000068G010 ..(KA's) i ANSWER: 056 (1.00) d.
[+1.0) i i REFERENCE: 1.
2203.005, " Loss of Containraent Integrity", page 1.
,l 2.
KA 000069A201 (3.7/4.3)
- .
000069A201 ..(KA's) i ' i
- ..
.... . .,, ~. - ~ - ~ ....... -. -..... ... -. -. -...... - ~..... ~. -. ... . - - ' SEWIOR REACT SPERATOR Page 79
i ' ANSWER: 057 (1.00) .c. (+1.0) , !
- REFERENCE:
. " Heat Removal Decision Tree", page 1.
1.
2202.009, 2.
KA 000074G011 (4. 5/4. 6) > 000074G011 ..(KA's) t ' ANSWER: 058 (1.00) c.
{+1.0)
REFERENCE: 1. - Technical Specification Section B3.4.8, " Specific Activity", page B 3/4 l 2.
KA bOOO76K306 (3.2/3.8),'000076G004 (2.1/3. 7). [ t 000076G004 000076K306 .(KA's) . ANSWER: 059 (1.00) i b.
(+1.0] -; ' - REFERENCE:
1, 2203.003, CEA Malfunction", page 2.
! " 2.
AA-S2002-012, " Control Element Drive Control", Terminal' Objective, page 2.
3.
KA 000001G010 (3.9/4.0)
000001G010 ..(KA's) . r . ANSWER: 060 (1.00) l
'd.
[41.0]- - , .. P epq g V uw s y
. - _ - .- t [ SENIOR REACTOR OPERATOR Page 80 REFERENCE: 1.
2203.003, "CEA-Malfunction", page 2.
2.- 'AA-52002-012, " Control Element Drive Control", Terminal Objective, page- _2. 3.
KA 000003G011 ( 4. 0/4.1) 000003G011 ..(KA's) ANSWER: 061 (1.00) b.
[+1.0) REFERENCE: 1.
2202.009, " Functional Recovery Procedure", page 11.
2.
KA 000007G012 (3.8/3.9) -000007G012 ..(KA's) ANSWER: 062 (1.00) .d.
[+1.0) REFERENCE: 1.
AA52002-015, " Loss of Coolant' Accident EOP", Objective 15 4, page 17.
2.
KA 000008G011 (4.0/4.1) 000008G011 ..(KA's) ANSWER: 063 (1.00) -- c.
[+1.01 ' REFERENCE: 1.
AA52003-015, " Loss of Coolant Accident", Objective:15.3, page 15.
2.
KA 000009K101 (4.2/4.7) 000009K101 ..(KA's) . _ _ _ _ _ _
- - s.
L.
1 SENIOR. REACTOR OPERATOR Page 81 ! ., ANSWER: 064 (1.00) ! a.
[ + 1. 0 ]- l ( REFERENCE: 1.
2202.003, " Loss Of Coolant Accident, Safety Function Status Check", page--i 72.
2.
KA 000009A239 (4.3/4.7) ' ' , 000009A239 ..(KA's)
ANSWER: 065 (1.00)
a.
[+1.0] REFERENCE:
1.
'AA52003-015, " Loss of Coolant Accident", Objective 15.3, page 16.. 2.
KA 000011K313 ( 3. 8 /4 '. 2 ) 000011K313 ..(KA's) . ANSWER: 066 (1.00) c.
[+1.0)
REFERENCE: 1.
2202.019,'" Standard Attachments", Exhibits.2 &;3, pages 90 & 91.
2.
AA52002-015, " Loss Of Coolant Accident EOP", Objective 15.15, page 42'. 3.
KA 000011A210 (4. 5/4. 7) 000011A210 ..(KA's)
- ANSWER:
067 (1.00) 'c.
(+1.0) .] l .. 's I
- r . - -,... -
- ,.
.+ a.
-.
-. -. -... -.. - - -., >. . G [S'EN10R REACTOR OPERATOR Page 82-- ~ REFERENCE: '1.
2202.001, " Standard Post' Trip Actions", page 8, 2.- 2202.010, " Standard Attachments", Attachment 1 and 2, pages 3 andi4.
3.
KA 000011G010 -(4. 5/4. 5),. 000011A103 (4.0/4.0) > 000011A103 000011G010 ..(KA's) , , . ANSWER: 068 (1.00) .a.
[+1.0) REFERENCE: l '. 2202.003, " Loss Of Coolant Accidents", page 2.
2.
KA 000011G007. (3.7/3.9) 000011G007 ..(KA's) ANSWER: 069 (1.00)
d.- .[41.0) REFERENCE: 1.
2203.029, " Loss - of Shutdown Cooling, page 3.
~ 2.
KA 000025A112 (3.6/3.5) . 000025A112 ..(KA's) - ANSWER: 070 (1.00) a.
[+1.0] I . -REFERENCE:
1.
2203.029, " Loss of Shutdown Cooling", page 1.
. . ] 2.
KA 000025A102 (3. 8 /3.9 ), 000025A207 (3.4/3.7), 000025A101-(3.6/3.7)
-000025A101 000025A207 000025A102- ..(KA's) l i .' . - . . . . , .-. - - --
' SENIOR REACTORLOPERATOR Page 03 ANSWER: 071-(1. 0 0 ) c.
(+1.0) REFERENCE: 1.
STM-2-02, " Control Element Drive Control", page 4.
2.
KA 000029K312 '(4.4/4.7) 000029K312 ..(KA's) -ANSWER: 072 (1.00) . a.
[+1.0) REFERENCE: 1.
Unit 1 Technical Specifications 3.9.2, " Refueling Operations",- page 3/4
l 9-2.
l 2.
KA 000032G000 (2.8/3.3) ' 000032G008 ..(KA's) ANSWER: 073 (1.00) c.- [+1.0]' REFERENCE: 1.
2203.038, " Primary To Secondary Tube Leakage", page 3.
2.
KA 000037A212 (3.3/4.1) 000037A212 ..(KA's) , ANSWER: 074 (1.00) a.
[+1.0) - . . . --. .- .
., -.~..~.w.
. -... . -. .. . -. ~ .... - . --. ... ~ - -.. .... '[ SENIOR'REACTOROPERATOR' Page 84 - A+ ' REFERENCE: h 2'203.038, " Primary To Secondary Leakage", page 3.
2.
AA52003-008,." Steam Generator. Tube Rupture", Objective 08.2, page 8.
3.- KA'000037A206 (4.3/4.5) 000037A206 ..(KA's)
ANSWER: 075 (1.00) , b.
[+1.0) .. REFERENCE: '1.
AA52003-008, " Steam Generator Tube Rupture", Objective 08.8, pages 19-20.
2.
KA 000038G011 (4.2/4.3) , 000038G011 ..(KA's)
, ANSWER: 076 (1.00) d.
[ +1. 0]. REFERENCE: 1.
2202.004, Steam Generator Tube. Rupture", page 30.
" 2.
AA52003-008, " Steam' Generator Tube Rupture", Objective 08.12,'page 28.
3.
KA 000038A217 (3.8/4.4) 000038A217 ..(KA's) ANSWER: 077 (1.00) ' d.
-{ + 1. 0) i REFERENCE: 1.
AA52003-010, " Loss Of Feedwater", pages 8 & 15.
2.
2202.006, Loss Of Feedwater".
" !, 3.
KA 000054A203 (4.2/4. 3 ), 000054K102 (3.6/4.2) 000054K102 000054 A203 ..(KA's)- _ . f p e.,.asy --.4,, ur,, .~w- + ,n g ye- , . .
-- - -. . _.. SENIOR' REACTOR OPERATOR Page.85 ' ANSWER: 078 (1.00) -a.
[+1.0) ' REFERENCE: 1.
2202.006, " Loss of Feedwater", page 2.
2.
AA52003 010, " loss Of Feedwater", page 11.
i.
3.
KA 000054K304 (4.4/4.6) , ' 000054K304 ..(KA's) , , ANSWER: 079 (1.00) , b.
[+1.0) ' REFERENCE: .; 1.
AA52002-007, " Electrical Distribution", page 32.
' 2.
KA 000058A203 (3.5/3.9) 000058A203 ..(KA's) f ^ ANSWER: 080 (1.00) c.
{+1.0] REFERENCE: 1.
AA42102-002, " Pressurizer Pressure and Level Control System", Objective 5, pages 43-45.
2.
KA 000028A212 (3.1/3.5) e t 000028A212 ..(KA's) ANSWER: 081 (1.00) b.
[+1.0)
. ,, , _.. _ _ . _
, _ .-. _ _ _.. - -... -. - _ _. _ _ _,____._- .. _ ,. -, _.. _.,_.....m._..,, - i i-SENIOR REACTOR OPERATOR-Page 86~ ~ TREFERENCE: 1.
.2502.001,." Refueling Shuffle", Attachment J, page 30.
-2.
KA 000036G010 (3.7/3.8) 000036G010 ..(KA's) S l ANSWER: 082 (1.00) a.
(+1.0] e . -REFERENCE: 1.
AA52003-011, " Loss of Offsite Power", Objective 11.6, page 21.
2.
KA 000056A288 ( 4.1/ 4. 2 ) 000056A288 ..(KA's)'
' ANSWER: 083 (1.00) ' a.- -[+1.0)
REFERENCE: - 1, 2203.021, " Loss Of Instrument Air", Attachmen't E, page 41'. -2.
KA 000065A208 (2.9/3.3) > - . 000065A208 ..(KA's) s l ANSWER: 084 (1.00) " .a.
. [ + 1.' 0 ] '
REFERENCE:: ' .1. l'012.019, " Radiological Work Permits",-page 6.
- i ' .2.
KA 194001K104 (3.3/3.5), 194000K103 (2.8/3.4) - . . .1 - 194000K103 194001K104 .. '( KA ' s ) ] - 1
..l
- l " . i . ] .. , ,,,a ., -n - n.
,,, . . -- - - - - n -.
. -. ..s.
.. -,. - . .... .... . . .. - ... .. . .... i P LSENIOR' REACTOR OPERATOR Page 87- ~ . , ~ ANSWER: 085 (1.00) c; [+1.0] L REFERENCE: 1.
1053.005, " Confined Space Entry Program", page 4.
2.
KA 194001K111 (3.4/3.5) , 194001K111 (KA's) .. ANSWER: 086 (1.00) d.- [+1.0) , REFERENCE: '1.
1053.005, " Confined Space Entry Program", page 9.
2.
KA 194001K114 (3.3/3.6) 194001K114 (KA's)
.. , ANSWER: 087 (1.00) d.
[+1.0) ' ' REFERENCE: 1.
1053.005, " Confined Space Entry Program", page 13.
2.
KA 194001K114 (3.3/3.6) , 194001K114 .(KA's) . . ANSWER: 088 (1.00) d.
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. . SENIOR REACTOR' OPERATOR Page 88
REFERENCE: 1. - 1000.106, " Primary Chemistry Monitoring Program", page 4.
-2, KA 194001A114 (2.5/2.9) 194001A114 ..(KA's) -ANSWER: 089 (1.00) c.
[+1.0] ' REFERENCE: 1.
T/S BASES 3/4.4.7 2.
KA 194001A114 (2.5/2.9) 194001A114 ..(KA's) . . t-ANSWER: 090 (1.00) c.
[+1.0) x ' i REFERENCE: 1.
1000.128, " Industrial Safety and Occupational Health", page 46.
2.
KA 194001K107 (3.6/3.7) 194001K107 ..(KA's) ANSWER: 091 (1.00) , '- a.
[+1,0) REFERENCE: 1.
2202.002, " Reactor Trip Recovery", page 2, 2.
KA 194001A102 ( 4.1/ 3. 9 ) 194001A102 ..(KA's) -1 l
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. _ _ _ _ _, _.. m. _. ._ .. - _.. . _ _. - . - . . - , SENIOR REACTOR' OPERATOR Page 89 A PANSWER: 092 (1.00) i d.
[+1.0] . REFERENCE: 1.
1062.001, " bud: Reporting", page 24.
' 2.
KA 194001A116 (3.1/4.4) 194001A116 ..(KA's) ANSWER: 093 (1.00) c.
[+1.0] REFERENCE: ' 1.
1000.027, " Hold and Caution Card Control", page 31.
2.
KA 194001K102 (3.7/4.1) 194001K102 ..(KA's) ANSWER: 094 (1.00) d.
[+1.01 REFERENCE: 1.
1000.027, " Hold and Caution Card Control", page 39.
2.
KA 194001K102 (3.7/4.1) , 194001K102 ..(KA's) ANSWER: 095 (1.00) c.
[+1.0]
- 1
.l ' .. -, _.- __
. . -. . . - .. ... SENIOR' REACTOR OPERATOR ' Page 90 - REFERENCE: + 1.
1000.006, " Procedure Control", page 5.
2.
KA'194001A101 ( 3. 3 / 3. 4 ') ' 194001A101 ..(KA's) l ! ANSWER: 096 (1.00) . a.
[+1.0] ' r REFERENCE: 1.
1015.001, Conduct o2 Operations", Attachment 1, page 56.
~ " 2.
KA 194001A103 (2.5/3.4) . L 194001A103 ..(KA's) ! ' , . ANSWER: 097 (1.00) d.
{+1.0] j REFERENCE: l 1.
1015.032, " Operations-Procedure Users Guide page-6.
. , 2.
KA 194001A101 ( 3 '. 3 / 3. 4 ) - 194001A101 ..(KA's) , ANSWER: 098 (1.00) b.
(+1.0] ' REFERENCE: 1.
1015.005, " Operations Key Control", page 5.
2.
KA 194001K105 (3.1/3.4) 194001K105 ..(KA's) , , N N-n v-w.
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[+1.0) l ' REFERENCE: 1.
1015.001, " Conduct Of Operations", page 36.
2.
KA 194001A103- (2. 5/3.4) - ' 194001A103 ..(KA's)
l ' ' ANSWER: _100 (1.00) d.
[+1.0) . -
REFERENCE: , ' 1.
1015.007, " Fire Brigade Organization and Responsibilities", page.5.
2.
KA 194001K116 (3.5/4.2) - 194001K116 ..(KA's)
,
, I b b ('********* END OF EXAMINATION **********)
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f ANSWER KEY , t ! -.! MULTIPLE CHOICE 023 a 001 c 024 d ' 002 b 025 c , 003 b 026 c , 004 a 027 d - 005 d 028 c . .! 006 d 029 b [ 007.
c 030 a y ~000 c 031 a.
j 009 d 032 d . 010 c 033 d 011' t 034 b.
, 012 c 035 d 013 c 036 c . .014 d 037 b '015 a 038 d-016 c 039 b 017-d 040 d > 1018 a 041 d' 019 b 042 a 020 a 043 c 021 c 044 a . '022 a 045 d' . ,
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AN'SWER KEY
!
046 c 069 d 047 b 070 a I 048 a 071 c 049' d 072 a 050 b 073 c 051-c 074 a
' 052 a 075 b 053 d 076 d 054 b 077 d 055 b 078 a " 056-d 079 b , 057 c 080 c , 058-c 081 b 059 b - 082 a 060 d 083 . a - , 061 b 084 a 062 d-085 c 063 c 086 d 064 a 087 d 065 a' 1088 d
066~ c 089 c.
067 Lc 090 c 068 a 091 a - v.. e-eu= -w n ,4 g-
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ANSWER KEY 092 d 093 c 094 d 095 c 096 a 097 d 098 b 099 c 200 d +
i (********** END OF EXAMINATION **********)
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