IR 05000424/1992302
| ML20127G700 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 01/05/1993 |
| From: | Lawyer L, Mcwhorter R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20127G684 | List: |
| References | |
| 50-424-92-302, NUDOCS 9301220045 | |
| Download: ML20127G700 (213) | |
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. NUCLEAR REGULATORY COMMISSION REGION tl $ / ,l $ l~ 101 MARIETTA STREET. * ?%. [4 ATLANTA, GEORGI A 30323 9***+'t/- ENCLOSVRE 1 EXAMINATION REPORT - 50-424/92-302 Facility Licensee: Georgia Power Company Facility Name: Vogtle Electric Generating Plant Facility Docket Nos.: 50-424 and 50-425
_ Facility License Nos.: NPF-68 and NPF-81 The NRC administered examinations at the Vogtle Electric Generating Plant near Waynesboro, Georgi ^ venJ.,,* e' &,- 7"?4?c4 sen/ /-m f';6L , Al Chief Examiner: / - 5 - 13 Richard D. McWhorter, J L/ Date Signed Examiners: T. Guilfoil, Sonalysts, In G. Weale, Sonalysts, In Approved By: #$ b 4 Lawren(6 L. Litwyer, Chief 9 7 - 73 Date Signed Operator Licensing Section 1 Division of Reactor Safety SUMMARY , Scope: Regular, announced initial-examinations were conducted during the weeks of December 7 and 14, 1992. The examinations were administered under the guidelines of the Examiner Standards, NUREG-1021, Revision'6. Initial written and operating examinations were administered to four Senior Reactor Operator and five Reactor Operator candidates. A requalification retake examination was administered on December 3, 1992, for one Senior Reactor Operator who failed the requalification examination given in June 199 Results: Four Senior Reactor Operator and five Reactor Operator candidates passed.the initial examination. One Senior Reactor Operator passed a requalification retake examinatio PDR ADOCK 05000424 V PDR _ _ _ _ _ _ _ _ _ _ __ t
_ _ _ __ -__ REPORl DETAILS Persons Contacted Licensee Employees
*J. Beasley, Assistant General Manager
*R. Brown, Supervisor, Operations Training
*C. Burke, Specialist, Safety Audit and Engineering Review
*R. Doorman, Manager, Training and Emergency Preparedness
* Gabbard, Specialist, Technical Support G. Gunn, Instructor, Operations Training
*C. Meyer, Operations Superintendent
*T. Mozingo, Oglethorpe Power Corporation C. Salter, Instructor, Operations Training -
* Sheibani, Supervisor, Nuclear Safety and Compliance
* Shipman, General Manager S. Wilkerson, Instructor, Operations Training
*J. Williams, Unit 1 Operations Superintendent NRC Employees B. Bonser, Senior Resident Inspector
*P. Balmain, Resident Inspector
*J. Starefos, Resident inspector
*D. Starkey, Resident inspector
* Attended exit meeting Other licensee employees contacted included instructors, operators, engineers, and office personne . Discussion Results The examination was administered under the guidelines of the Examiner StTndards, NUREG-1021, Revision 6. Four Senior Reactor Operator (SRO) and five Reactor Operator (RO) candidates passed the examination. One SR0 passed a requalification retake examinatio Reference Material The licensee supplied reference material to support examination preparation. Procedures, lesson plans, and other documents provided were adequat Examination Development Representatives from the licensee training staff met with examiners in the regional office on November 23 and 24,1992, to review the proposed examinations. The licensee's input to the initial examina-tion development process was valuable and increased the accuracy and quality of information contained in the examinatio _ _ _ _ . . _ _ _ . _ - . _ _ _ _ _ _ , . . _ _ _ _ _ _ _ ____ _ _ _
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Report Details 2 During the week of November 30, 1992, the examiners = worked with licensee personnel on site to prepare the examination. Simulator scenarios and Job Performance Measures (JPMs) were validated in the-simulator and the plant. Additionally, final comments to the written examination were resolve The licensee's support to this i effort was very constructiv The licensee also supplied the Nuclear Regulatory Commission (NRC) with a proposed examination for the requalification retake examina-tion. The examination was good in content and structure, and met the guidance of NUREG-1021 with regards to format for retake exam-inations. The NRC reviewed the examination and proposed one sub-stitute JPM that was incorporated into the examint:av after licensee revie d. Examination Administration The written and simulator portions of the examination were adminis-tered without major problems. Licensee support in operating the simulator and providing other assistance was goo , During the administration of the walk-through portion of the examination, a problem occurred in that an excessive amount of time was taken by several candidates. One SR0 candidate took ten hours, and one R0 candidate took nine hours. Three additional candidates took seven hours. The length of these walkthroughs was signifi-cantly in excess of the averages for candidates at other facilitie The primary reason for the abnormally long walkthroughs was the
- excessive use of references by the. candidates. In several cases, examiners concluded that the need for the excessive use of refer-ences was related specifically to below average knowledge, and was
! not solely attributable to stress or to candidates substantiating l answers to questions. This below average knowledge level demon-i strated by several chndidates who required'an excessive use of references on the walk-through portion of the examination is identified as a-weakness in the engineering and technical support functional are Long walk-through examinations were to a lesser extent. also attributable to .the aggressive use of active simulator JPMs by the-NR The examiners agreed with the licensee to-use lessons learned during future' examinations to attempt to streamline the scheduling of walkthrough The licensee provided an evaluator to assist in the administration of the requalification retake' examination. No problems were experienced with the administration of this portion of the-examinatio ,
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Report Details 3 Candidate Performance As evidenced by the high percentage of candidates passing the examination, candidate performance was generally good, except as noted in paragraph 2.d above. During the simulator portion of the examination, cannidates demonstrated an excellent ability to use emergency procedures successfully to mitigate the effects of simulated complex accident situations. The examiners concluded that this reflected an appropriately strong emphasis on the training of candidates in using emergency procedures in the simulator. The excellent performance of candidates during the simulatsr portion of _ the examination is identified as a strength in the engineering and technical support functional are Several candidates demonstrated similar minor performance deficiencies in three area First, several candidates did not know from memory Technical Specification (TS) Limiting Condition for Operation action items with a time of one hour or less (i.e.: 3. action a, 3.1.1.4 action, and/or 3.1.3.1 action b). Second, most candidates did not know how to locate and read Nuclear Service Cooling Water supply temperature as required by Abnormal Operating Procedure 18021-C, " Loss of Nuclear Service Cooling Water," step Third, several candidates appeared confused and unnecessarily delayed manually initiating a containment ventilation isolation when a valid initiating signal was presen Additionally, it was observed that several operators could only locate important information in TS or procedures by turning through documents page by page, thus taking an unreasonable amount of tim Examiners concluded that this reflected a lack of candidate - familiarity with reference materia Simulator Fidelity Three minor problems in simulator fidelity were observed and are listed in Enclosure 2. The first two problems were corrected prior to the completion of the examinatio Resolution of the third was ongoing at the end of the examinatio Additionally, the simulator was not capable of simulating several malfunctions that would have been useful for examination purposes to reflect events that actually have occurred in nuclear power plant Items that examiners found the simulator unable to simulate included: (1) loss of power to the switchyard, (2) loss of individual offsite power lines, (3) failure of Refueling Water Storage Tank (RWST) level transmitters, (4) failure of a single Main Steam Isolation Valve (MSIV) to close, (5) failure of a Solid State Protection System (SSPS) logic output, and (6) loss of annunciator _ _ _ - _ _ _ _ _ _ _ _ _ _ _
Report Details 4 9 Procedures Four minor problems were identified with procedures used during the examinatio First, while evaluating a simulator fidelity issue associated with Steam Generator wide range level instruments, the licensee identified that A0P-18032-1, " Loss of 120V AC Instrument Power," Attachment F, incorrectly listed LI-504 to lose power on a loss of IDYlB. Second, S0P-13501-C, " Nuclear Instrumentation System," step 4.1.3.2, required setting the local meter gain potentiometer to a value of 0.1. However, when performing the procedure at power, that setting drove the local meter offscale - hig Third, SOP-14701-1, " Reactor Trip Breakers UV and Shunt Trip Test," step 5.1.6.d stated to rack in bypass breaker A to the " operate" position, but the position indicator listed the fully racked in position as " connect." Finally, E0P-19102-C, "ECA-0.2 Loss of All AC Power Recovery with SI Required," step 3.b response not obtained, was inconsistent in the format for identifying valve numbers, unit, c and location with the rest of the procedur Additionally, examiners reviewed the status of resolution for four procedure problems identified in NRC Examination Reports 50-424/91-302 and 50-424/92-301. The examiners found that the licensee had corrected problems identified with procedures A0P-18007-C, A0P-18031-C, and S0P-14980-C. A fourth problem identified with the format of procedure A0P-18003-C had been resolved by the licensee not to require a procedure change. The licensee had assessed the format and concluded that the current procedure was consistent with procedure writing guides. Examiners - reviewed this action and considered it adequate to resolve the issu Material Condition of the Plant During examination activities associated with the diesel fire pumps, examiners noted that the nomenclature used for valves on the pumps was confusing. The cooling water valves for the diesel fire pumps were labeled for fire pump numbers 3 and However, common plant terminology and other valve labels in the area, referred to the diesel fire pumps as pump numbers 1 and 2. The licensee stated that the problem had been previously identified, and that the valves were labeled in accordance with plant drawings. However, the drawings for the different systems involved used different numbering schemes for the pumps. The licensee was continuing to evaluate the issu Within the areas inspected, no violations or deviations were identified.
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_ . _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . _ _ . _ __ _ _ _ _ _ _ _ _ Report Details 5 Action on Previous Inspection Findings (Closed) IFl 50-424; 425/92-301-02, Unauthorized Operator Aids Exist in Plant on Systems Utilized in JPM Performance. This item concerned the fact that during a previous NRC examination, it was noted that various component identification numbers were written on the walls, support plates, pipes, access doors, and ladders using pencils, magic markers, etc. During this examination, examiners made several tours throughout accessible areas of the plant looking for unauthorized operator aid Examiners found that the licensee had removed previously identified unauthorized markings, and no new problems were identified. Examiners - considered the licensee's corrective action to be adequate, and this inspector follow-up item is close . Exit Interview At the conclusion of the site visit, the examiners met with represent-atives of the plant staff listed in paragraph I to discuss the results of the examinations and inspection findings. The licensee did not identify as proprietary any material provided to, or reviewed by, the eraminer / _ Y 'Y _ _ _ _ _ _ _ _ _ _ . . _ _ _ _ _ _ _ _ _ _ _ _
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ENCLOSURE 2 SIMULATOR FIDELITY REPORT , l Facility Licensee: Georgia Power Company Facility Docket Nos.: 50-424 and 50-425 Operating Tests Administered On: December 3 and 14 - 17, 1992 This form is used to report observations. These observations do not constitute, in and of themselves, 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 that may be used in future evaluations. No licensee action is required solely in response to these observation During the conduct of the simulator portion of the operating tests, the following simulator fidelity items were observed: ITEM DESCRIPTION
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SG WR Level LI-503 failed low upon a loss of power to ICY 1 Instruments LI-504 did not fail low on a loss of power to ICYl * FHB Isolation Upon resetting a Fuel Handling Building Reset ventilation isolation at panel QHVC,-the reset lights did not ligh N-42 Drawer Power Upon -a high -failure of the : lower detector to Meter power range nuclear instrument N-42,-the power meter on the instrument drawer-indicated -100 percent instead of off-scale hig * Fidelity problems corrected prior to the completion of the examinatio Additionally, the simulator was not capable of simulating several malfunctions that would have been useful for examination purposes and reflect events that
.actually have occurred in nuclear power plants. Items that examiners found the simulator unable to simulate included: (1)_ loss of power to the switchyard,_-(2) loss of individual offsite power lines,_ (3) failure of RWST_
level transmitter.s, (4) failure of a -single MSIV to close, (5) failure _ of an SSPS logic output, and (6) loss of annunciator . -- . .. . _
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U. S. NUCLEAR REGULATORY-COMMISSION- .,
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SITE SPECIFIC' E7MINATION.' I REACTOR OPERATOR LICENSE-REGION- 2 CANDIDATE'S NAME: _ FACILITY: Vogtle 1 &2 REACTOR TYPE: PWR-WEC4 _ _ . -i DATE ADMINISTERED: 92/12/10 INSTRUCTIONS TO CANDIDATE: Use the answer-sheets provided to document your answers. Staple this cover = sheet on top of the answer sheet Points-for each question are-indicated in: parentheses after the questio The passing grade requires a final grade of- ' at least 80%. Examination papers will be picked-up four (4) hours after the-examination starts, s
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CANDIDATE'S TEST VALU SCORE _
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100.00 % FINAL-GRADE , All work done on this examination is my ow I have neither given nor-received aid, f
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-REACTOR OPERATOR- O o Peet 2 1 ANSWER S.H E E-T ,
-j-1 Name: -i I
Multiple-Choice- (Circle or X your choice) l
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i-If you change your answer, write your selection in the blank, - l 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
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-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 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 d ' 013- a b c d 036- 'a .b c- d U14 - a b .c d 037 a li c Tl 015 & b c d 038 a b c dL
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 020 b c d- 043 a .c- d 021 a b c d 044 -a _c .d 02 a b c d 045 a- b c d !
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REACTOR OPERATOR (J Page 3 ANSWER SHEET Name: 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 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 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
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.A-N S W E S'H-E E T-Name:
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Multiple Choice (Circle or X your choice) l If you change your answer, write your selection ~in the olan 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 s
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O 0: Pese 4 c 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 penaltie . 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 examinatio . Restroom trips are to be limited and only one applicant at a time may leav You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.
' Use black ink or dark pencil ONLY to facilitate legible reproduction . Print your name in the blank provided in the upper right-hand corner of the examination cover sheet and each answer shee . Mark your answers on the answer sheet provide USE ONLY THE PAPER PROVIDED AND DO NOT WRITE ON THE BACK SIDE OF THE PAG . Before you turn in your examination, consecutively number each answer ' sheet, including any additional pages inserted when writing your answers on the examination question pag . Use abbreviations only if they a-e commonly used in facility literatur Avoid using symbols such as < or > signs to avoid a simple transposition.
, error resulting in an incorrect answer. Write it ou . The point value for each question is indicated in parentheses after the g questio . Show all calculations, methods, or assumptions used to obtain an answer to , any short answer questions.
l-l- -11. Partial credit may be given except on multiple choice questions. .Therefore,
ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLAN . 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 L points, and you give five responses, each of your responses will be worth j 0.20 point If'one of your five responses is incorrect, 0.20 will be
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l- deducted and your total credit for that question will be 0.80 instead of ' 1.00 even though you got the four correct answers.
L 13. If the intent of a question is unclear, ask- questions of the examiner onl _ g . .. w--m .-i tp l
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l 14. When turning.in_your; examination,-assemble'the completed examination.with examination questions; examination aids.and answer sheets. -In addition,
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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 examinatio . To pass the examination, you must achieve a grade of.'80% or greate . There is a time limit of four (4) hours for completion of the examinatio . 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 (3.00). - Unless prior exception approval has been obtained, when must.a-
. controlled key issued for -use in the protected area _ (PA) be returned?
a. At shift en . b. Daily by 083 ' c. No time limit providing the key is maintained in the P d. No time limit providing the key is signed out to an individua "
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' QUESTION: 002 (1.00)
When restoring a FAIL CLOSEfair-operated valve (AOV) that ..as-a handwheel, after it has been used as a closed boundary valve for a clearance,~which one of the following correctly describes the order (first to last) for removing the hold tags?- a. Handwheel, valve air supply, handswitch , b. Handswitch, handwheel, valve-air supply c. Valve air supply, handswitch, handwheel d. Handwheel, handswitch, valve air supply
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L QUESTION: 003' (1.00) Following repair-of a-damaged breaker, which clearance status is required for breaker testing to be performed?. a. All hold tags must be temporarily cleared, the breaker racked to the test position, and a hold tag placed on the racking device, b. Necessary hold tags must remain on the breaker,-the breaker racked to the test position, and a caution tag placed on the racking devic c. All hold tags must bi removed and replaced with caution' tags,
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and the breaker racked to the test positio d. Necessary hold tags must be removed via a functional release, and the breaker racked to the test positio ! QUESTION: 004 (1.00) Which one of the following describes how an independent verification of the position of a manually operated THROTTLE valve is' performed? a. Compare a visual c';servation of the stem position or indicator
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position with the required positio b. Completely close the valve and then open it to the required positio ' c. Completely open the valve and then close it to the required position, d. Move the valve exactly one turn in the closed direction and then - return it one turn to its required positio !
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REACTOR-OPERATOR- _ LPage_- 9
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QUESTION:--005 (1. 0 0 ) '
; -.The1USS must authorize surveillance tests by signature if - the test::
a.-affects both unit , b. manipulates plant equipmen c. is' required for a mode change, d. has less than a 72-hour Action Statement,
~ QUESTION: 006 (1.00)
Per VEGP 10000-C, " Conduct of Operations," which one_of_the following may NOT be adjusted by shift operating personnel during at-power-- operations?. a. Power range gain adjustment following a calorimetri b. Intermediate range compensating voltage settin c. Data A/B selector switch behind the DRPI d.d splay, d. Atmospheric relief valve setpoint.
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-REACTOR. OPERATOR' -- Page l'O QUESTION: 007- (l'00)- .
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.You'are the Unit 1 RO with the unit in Mode 1. Th'e BOP operator is in- 7
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the Clearance and-Tagging Office obtaining an. aperture: car Which one of the following describe 6 the minimum relief requirements'for you to leave the control room for 45 minutes for a Fitness for Duty test? a. Conduct a joint walkdown of the control board with a licensed reactor operato b. Conduct a short term relief turnover in accordance with VEGP 10004-C, " Shift Relief," with another licensed reactor operator on shif c. Have the USS assume the combined watch and get permission from the SS to leave for the test, Conduct a full shift turnover to another licensed reactor operato QUESTION: 008 (1.00) ' Which one of the fellowing describes the lock reinstallation l requirements for a locked valve with a remote operator (reach rod) after l- the valve has been independently verified in its locked position? a. The lock must be reinstalled on the remote operator-(with a Caution tag on the local operator).
b. The lock must be reinstalled on the local operator.(with a l Caution tag on the remote operator).
c. The locks must be reinstalled on both the local and remote operators, and the reinstallations must be independently-verified.
l d. The lock must be reinstalled on the remote operator, and the-reinstallation must be independently verified.
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.-REACTOR ' OPERATOR O_ rese 11: 1
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Which one:of _the following combinations describes "the extremities" wtunt considering radiation-dose limits? a. Hands and feet only-b. Hands, arms,-ankles, and-feet ! c. Hands, arms, legs, and feet , d. Hands, forearms, ankles, and feet QUESTION: 010 (1.00) Emergency exposure limits are (1) rem whole body for life saving
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activities and (2) rem whole body exposure for corrective or protective activities, (1) 50; -(2) 10 (1) 50; (2) 25 - c. (1) 75; (2) 10 d. (1) 75; (2) 25 QUESTION: 011- (1.00) In accordance with ALARA guidelines, Independent Verification'of
, components is not-required if significant radiation exposure is involved. Which one of the following is the LOWEST whole-body. dose-that would constitute significant radiation exposure?
a, 5 mrem mrem
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d. 25 mrem-
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REAC'IOR OPERATOR h h Page '12
- QUESTION: 012 (1.00)
Which one of-the following current whole-body exposures will cause a:"NO ENTRY" display on the RCA access point HP computer terminal upon an attempted log-in? Where applicable, assume an updated Form 4 IS on file, lower dose' limits - for pregnancy have NOT been chosen, and individual has NO exposure history other than that listed, a. Male temporary radiation worker with 50 mrem in current quarter,
, Female radiation wcrker with 350 mrem-in current quarter, c. Male radiation worker with 850 mrem in current quarte Female radiation worker w.th 4150 mrem in current-yea QUESTION: 013 (1.00)
An individual's TLD shall be processed before the individual is authorized to receive greater than: a. 1000 mrem in a quarte mrem in a quarter.
') c. 4000 mrem in a year, mrem in a year.
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_ _ _ _ _ - _ _ _ _ _ - _ _ - _ _ - _ _ - _ - _ _ _ _ _ - _ _ _ - _ _ _ _ _ _ - _ _ . REACTOR OPERATOR-0 0 vase 13 QUESTION: 314 (1.00) Given tne following conditions on-Unit 1:
- Reactor Power at 90% - Rod Control in Automatic --A' power transient causes a Tavg-Tref mismatch with Tref being degrees F greater than Tavg Which one of the following describes the controlling rod group speed and direction of motion? steps / min, In
_ steps / min, In c. 16 steps / min, Out d. 24 steps / min, Out-QUESTION: 015 (1.00) Which one of the following describes the operation of the Master Cycler in the Rod Control System? a. Selects group to be moved and sends "go" pulses to the slav cyclers in the correct sequenc b. Generates current sequence orders for power cabinets and detects-faults in its own and other circuit c. Selects control banks to be moved wilen in- manual or' auto and
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sends "go" pulses to the auxiliary cycler d. Generates current sequence orders for power cabinets and controls the rate at which banks are stepped.
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REACTOR. OPERATOR' Page 14 QUESTION: 016 ('1. 0 0 ) Given the following cor.ditions at Unit 1:
- Reactor Power at 10%, startup in progress - Turbine load being increased slowly -
b If rod control is inadvertently placed in AUTO under these conditions,- which one of the following describes the response of the control rods?
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a. The control rods will NOT move out until Tavg is at least degrees F greater than Tref, b. The control rods will move out at the rate of 8 steps per minute to keep Tavg and Tref within 1.5 degrees c. The control rods will NOT move until CB D is manually positioned to 12 step d. The control rods will NOT move out until Turbine Impuls Pressure (PT-505) exceeds 15% powe QUESTION: 017 (1.00) Which one of the following describes the parameter (s) which the RIL computer uses to generate a low rod position setpoint? a. Tavg and Tre b. Core delta- c. Tref and core delta- d. Tref and controlling group rod height.
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REACTOR OPERATOR f( ) lll Page 15 QUESTION: 018 (1.00) Which one of the following statements describes the basis for each , reactor coolant pump (RCP) having a Class 1E breaker in series with its i non-1E breaker? )
a. To provide electrical isolation from the associated 13.8 kv bus l during motor maintenance and circuit breaker testin b. To prevent a mechanical failure of the containment electrical l penetration due to fault current if the non-1E breaker fail c. To prevent damage to the supply breaker for the associated l 13.8 kv bus in the event of an RCP locked roto ' I d. To prevent damage to the supply breaker for the associated 13.8 kv bus in the event of a seal failure grounding the motor.
QUESTION: 019 (1.00) Which one of the following parameters would eventually require stopping a Reactor Coolant Pump (RCP)? a. Upper motor radial bearing temperature - 180 degrees F, b. ACCW flow to Air / Water heat exchanger - 190 gp c. #1 seal leakoff flow - 6 gp d. Shaft vibration - 10 mils.
QUESTION: 020 (1.00) Which one of the following is the electrical bus which supplies power to Reactor Coolant Pump (RCP) 3 at Unit I? a. 1NA b. 1 NA c. 1AA0 d. 1BA0 REACTOR _ OPERATOk h- ( 'Page 16-QUESTION: 021 (1.00) ~ Unit.1 is operating at 100% power at BOL with all systems operable ~when-a. complete loss of instrument air occurs._ Which one'of'the_following_ lists Chemical and Volume Control System _(CVCS) valves:that will fail' OPEN? a. Charging Flow Control Valve (FV-121) ,_ RHR Cross _ Connect Valve (HV-128), b. Letdown Leak Protection Valve (HV-15214), Auxiliary Spray Valve (HV-8145).
c. Orifice Isolation Valves (HV-8149A, B, C), Containment Isolation Valves (HV-8160, 8152).
d. Seal Injection Flow Control Valve (HV-182), Letdown Pressure Control Valve (PV-131). , , QUESTION: 022 (1.00) Which one of the fo? lowing describes a starting limitation imposed upon the Residual Heat Removal (RHR) pumps? consecutive starts from ambient temperatur consecutive starts from operating temperature, start from cold conditions.
d. Subsequent starts permitted after 30 minutes if the motor is at a standstill.
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QUESTION: 023 (1.00) ! t
- During operation at 100% power at Unit 1, power is lost to the motor
' aprtrator on valve HV-8112, seal return line containment isolation valv , Which one of the following describes the effect on RCP seal return f parameters of the loss of power to HV-81127 l I ! a. Pressure will slowly rise until relief valve PSV-8121 open ;
b. Pressure will increase but the check valve around HV-8112 will !
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- open preventing relief valve actuation.
' c. Temperature will increase slowly due to reduced flow and may require RCPs to be stopped.
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d. Temperature and pressure will remain normal because valve HV-8112 will not change position, j QUESTION: 024 (1.00)
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L ' Which one of the following is the minimum required normal charging header flowrate if an Emergency Boration is to be performed from the Boric Acid Storage Tank (BAST)? _ a. 30 gpm
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b. 42 gpm c. 87 gpm , d. 100 gpm !
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REACTOR OPERATOR O O- vase 18 l
i i QUESTION: 025 (1.00) ~i Which one of the following will NOT result in the generation of a Containment Ventilation Isolation (CVI) signal on a high alarm? ; a. Containment area radiation low range monitor RE-0002, b. Containment vent particulate monitor RE-2565A . l, c. Containment vent iodine monitor RE-2565 ; d. Containment atmosphere particulate monitor RE-2562 ! QUESTION: 026 (1.00) i With the Solid State Protection System (SSPS) Train A Mode Selecto '
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Switch in " TEST", what is the condition of Train A SSPS? a. Input relays are prevented from providing a " tripped" input.to the logic car > b. The logic card is prevented from recognizing any blocks-or permissives that are active, c. The Spray Test and output Relay Test circuits are enable , d. Train A SSPS-is still able to actuate ESF equipment on a. valid ESF initiation signal,
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QUESTION: 027 (1.00) Which.one of the following. describes the' indication which would be obtained if a thermocouple circuit suffers a short in the junction box? a. Temperature indication would be off scale lo b. Temperature indication would be off scale hig c.-Temperature indication would oscillate at 1the-high end, d. Temperature indication would be that of the junction bo i .,a . . . _ . _ . . . . . . _ . . . . . . _ . . , . _ , . . . _ . _ . . _ . . . , , . , . . . . . .- . , . , . __ _ , - . - . , . . . . a,.,,,,,...,
. . . . . . . REACTOR OPERATOR () () Page 19 QUESTION: 028 (1.00) Which one of the following describes the response of the containment-cooling fans to a Safety Injection signal? a. Fans running in fast speed will stop; then all fans will start-in slow speed at 30.5 seconds, Fans running in fast speed will stop; then fans 3, 4, 5 and 6 will start in slow at 30.5 seconds, and fans 1, 2, 7 and 8 will start in slow after 50.5 second Fans running in fast speed will shift to. slow speed; then-the idle fans will sequence on in slow speed after_50.5 s3cond d. Fans running in fast speed will continue operating until-5 seconds;-then all fans will start / shift to slow spee QUESTION: 029 -(1.00) Given the following conditions at Unit 1:
- A reactor trip has just occurred from 25% power - No safety injection signals are present Tavg is 563 degrees F, decreasing slowly - All steam generator levels are greater than 27% narrow range Which one of the following describes the response of the main feedwater system?
a. Main feed regulating valves and bypass feed regulating valves close; MFPs trip and the main steam lines isolat b. Main feed regulating valves, bypass feed regulating valves, and bypass feed isolation valves close; and MFPs tri c. Main feed regulating valves and bypass feed regulating valves close; and the main steam lines isolat d. Main feed regulating valves, bypass-feed regulating valves, and bypass feed isolation valves clos _,
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REACTOR OPERATOR O O rese 20 QUESTION: 030 (1.00) Which one of the following describes the mechanism by which adequate flow is assured to intact steam generators (SGs) in the event of an AFW feedline break? a. Capacity of the AFW pump b. Sizing of the AFW feed lines, c. " Quick acting" isolation valves in AFW lines to SG d. Flow restrictors in AFW lines to SG _ QUESTION: 031 (1.00) Which one of the following describes the source of steam to the "trbine driven auxiliary feedwater pump? a. SGs 1 and 2 b. SGs 3 and 4 c. SGs 1 and 3 d. SGs 2 and 4 QUESTION: 032 (1.00) What is the lowest steam pressure at which the turbine-driven AFW pump can be operated? a. 30.psig, b. 60 psi c. 90 psig, d. 120 psi .
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QUESTION: 033 (1.00) i t Given the following conditions at Unit 1: I
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Reactor power at 35%
- Non-1E buses have been transferred to the UATs i - All buses are energized l -
No indicating lights displayed on high side circuit switcher for - the RAT Which one of the following describes the condition that would cause . these indications to exist? a. The RAT high side circuit switcher has auto trippe b. A fault exists in the RA c. Control power is lost, the RAT high side circuit switcher is close L; i I d. The fast bus transfer to the RAT has actuate QUESTION: 034 (1.00) Both a green light and a blue light are illuminated on the front panel of a standard Data Processing Module _(DPM) in the Digital Radiation-Monitoring System. Which one of the following describes the status of the DPM? * a. DPM power is on and a high radiation alarm' exists.
, b. DPM. power is in standby and1 trouble alarm has not been cleared, c. DPM has an alert or intermediate radiation alar , i d. DPM is in the bypass mode of operatio ,
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uw 0 : 035 (1. 00 ) l L.,un the-following conditions on Unit 2: r
- The reactor tripped from 100% power due to a trip of RCP l -
Pressurizer pressure control is in automati i
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RO fails to manually close loop 1 spray valv Which one of the following describes the response of the plant if RCS I pressure increases to 2260 psig? Loop 4 Pressurizer Spray Valve will reduce RCS pressure to 2235 psig,
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b. Pressure will be maintained at 2260 psig by the Variable Bank Heater , c. Pressure can ONLY be reduced manually by the operator using i Auxiliary Spra Pressure will maintain steady at 2260 psig until PORV PCV-455A opens to reduce pressure.
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r QUESTION: 036 (1.00) : Which one of the following describes the effect of decreasing the Pressurizer Pressure Master Controller potentiometer setting to 2185 psig during Mode 1 operation? a. Actual system pressure will remain between 2250 and 2260 psig-but the low pressure alarm will actuat b. Actual system pressure will remain in the normall control band of 2250-2260 psig, but the PORVs.will receive a block signa c. Actual system pressure will decrease to 2185 psig, and the PORVs will receive a block signa d. Actual system pressure will decrease to 2210 psig, and then the backup heaters will raise pressure to 2218 psig.
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REACTOR OPERATOR (]) () Page 23 QUESTION: 037 (1.00) Given the following conditions at Unit 1:
- Reactor Power at 100% - Pressurizer level control in automatic -
The CVCS positive displacement pump (PDP) is running with control in automatic
- A malfunction in the Auctioneering High Tavg circuit causes the output of auctioneered High Tavg to fail low Which one of the following describes the response of the system?
a. PDP speed will lcwer, actual pressurizer level will decrease to - 25% and stabiliz PDP speed will increase, actual pressurizer level will increase to 60% and stabilize, c. PDP speed will lower, actual pressurizer level will decrease to 17%, all heaters go off, reactor trips on low pressur PDP speed will increase, actual pressurizer level will increase to 92%, reactor trips on high pressurizer leve QUESTION: 038 (1.00) Which one of the following describes an accident which the ECCS is designed to mitigate? a. Loss of electrical powe b. High reactor power transient, High startup rate in the power rang Feedwater line brea $ _ _ _ _ _ . - . .
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~ QUESTION: 039 (1.00) 6 Which one of the following describes one or more penetrations in loop 3 cold leg, in addition to the accumulator / safety injection /RHR and BIT-conbuctions?
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a. CVCS letdown, b. Normal Charging, PZR spra ; i c. CVCS excess letdown, sample lin ' d. Alternate Charging, PZR spra QUESTION: 040 (1.00) Which one of the fo'. lowing describes the locations where reactor vessel j level (RVLIS) indi';ations are available? a. Plant Safe:y Monitoring System computer and the Emergency Response F acility (ERF) computer, b. OMC9 recorder and local display gages.- c. Proteus-computer and the Technical Support Center computer, d. QPCP level indications and ERP compute QUESTION:-041 (1.00) Which one of the following describes the effect of a loss of 125 VDC switchgear 1BD1? a. Power is lost to TDAFWP steam supply valve (HV-5106).
. . Power.is lost to RHR isolation valve from RCS-hot leg loop 1.- L
- Power is lost to 120 vac vital instrument bus-1BY1B, which' will -
autotransfer to the alternate power. suppl d. Power is lostEto 125 VDCffield_ flashing for ED/G_"B".
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QUESTION: 042 (1.00) The Control Room annunciator for Spent Fuel Pool High Temperature is out-
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of service. Which one of the following would be an alternate indication-of a loss of Spent Fuel Pool Cooling? , a. Local audible high temperature alarm in the Fuel Handling Buildin , I Local audible alarm of Spent Fuel Pool Cooling High Temperature on the Liquid Waste Processing Panel, c. Spent Fuel Pool High Humidity alarm in the Control Room, High-alarm on RE-008, Fuel Handling Building Area Radiation ;
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, QUESTION: 043 (1.00) Given the following conditions at Unit 1:
' - Refueling is in progress - A spent fuel element is being moved from the upender to the spent-fuel pool storage rack -
The spent fuel element is dropped to the bottom of the spent fuel pool, rupturing the element
Which one of the-following products, if-released from the= dropped / damaged spent fuel element, may cause automatic actuation of'the Fuel Handling Building Ventilation Isolation? a. Hydrogen ga b. Krypton and iodine gas, c. Alpha radiation from fission product d. Gamma radiation from corrosion products.
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I REACTOR OPERATOR O O case 26 l l l l QUESTION: 044 (1.00) Given the following conditions at Unit 1
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Reactor startup is in progress
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All NI switches are in their normal lineup
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No manual blocks have been inserted ;
- Intermediate Channel N35 indicates 2E 10 -
Intermediate Channel N36 indicates 9E-11
- Power is lost to Source Range Channel N31 j -
Power is maintained to Source Range Channel N32 Which one of the following describes the response of the reactor plant? a. A reactor trip signal is generated resulting in a reactor tri b. A reactor trip signal is generated but no trip occurs-since one channel is above P- ; c. No reactor trip signal is generated since one channel is above P d. No reactor trip signal is generated but the level trip switch ' must be taken to bypass as soon as N36 indicates greater than 1E-1 ;
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i QUESTION: 045 (1.00) i i Given the following conditions at Unit 1:
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Reactor Power ac 75% and steady ,
- Rod Position Indication Urgent Failure alarm actuated ? - ROD DEV/ RADIAL TILT alarm actuated - On the DRP1 the operator observes: , -
Rod H8 rod bottom light illuminated ;
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General Warning light illuminated ; Which one of the following describes the conditions of the plant? l a. Rod H8 has dropped due to a HOLD coil failur ; b. Central control card for Rod H8 has faile c. The control and display unit has faile d. Data A and Data B for Rod H8 have faile QUESTION: 046 (1.00) Given the following conditions at Unit 1:
- Reactor startup is in progress Source range-channel N31-indicates 1E5 - Source range channel N32 indicates 9.5E4 - Intermediate range channel N35 indicates 4E-10 -
Intermediate range channel N36 indicates 2E-11 Which one of the following statements describes the condition of.the nuclear. instruments?
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i REACTOR OPERATOR O O rase 28 l l QUESTION: 047 (1.00) )
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Given the following conditions at Unit 1:
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Plant is in Mode 3
- RCS pressure is 750 psig - RCS temperature is 400 degrees F , - Main steam line isolation signal has been actuated and not reset : - Steamline SI and Pressurizer SI have been blocked- l l
While attempting to reset the main steam line isolation signal, the operator mistakenly resets the low steamline pressure SI. Which one of the following describes the automatic response of the plant?. a. An SI signal will be generated and an SI will occ !
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b. An SI signal will be generated but, an SI will NOT occur since-the plant is in MODE An SI signal will NOT be generated since the plant is in MODE f d. An SI signal will NOT be generated since the plant-is-in MODE-3 and the MSLI is actuate L P
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REACTOR OPERATOR (]) lll QUESTION: C48 Page 29 (1. 0 0) Given the following con di tions at Unit 1:
- - Reactor in Mode 3 Preparations being nad e f or SSPS Train A Mode Select -
Input reactor startup
-- Trip Breaker A isError Inhibit Switch ior Switch is in TEST Trip Breaker B is closed s in INHIBIT If the operator racked out
, \ one of attempts to for testing li { the following describ es the response ofclose Bypass Breaker B fo The General Warning al the plant?r testing, which any reactor trip signalsarm for Train Bewill actuat
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b. A reactor trip signal close and inhibit The multiplexer will be generated when By pass Breaker B is returned to " Normal".willtoshift Train B until Logi Train B slave c Train A is relays will align for bypa ss breaker operations . QUESTION: 049 (1.00) Which one of System (7300) CABINETS PWR esult SUPPLY FAILthe following will r a. Loss of Process Cabinets?URE" annunciator for PCSthin illumination of t e Reactor Protection b. Loss the NYRS electrical us. b of 24 VDC backup power to N SSS protection test switches in TESTEnergir.ing the Master cabinet T d. Loss . est Relay before placing th of power to the Eagle 21 e bistacle cabine . , e
REACTOR OPERATOR (]) (]) Page 29 l QUESTION: 048 (1.00) Given the following conditions at Unit 1:
- Reactor in Mode 3 l -
Preparations being made for reactor startup t
- SSPS Train A Mode Selector Switch is in TEST ' -
Input Error Inhibit Switch is in INHIBIT .
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Trip Breaker A is closed
- Trip Breaker B is racked out for testing i If the operator attempts to close Bypass Breaker B for testing, which one of the following describes the response-of the plant?. ,
a. The General Warning alarm for Train B will actuate and inhibit' any reactor trip signal b. A reactor trip signal will be generated when Bypass Breaker B is close c. The multiplexer will shift to Train B until-Logic Train A is returned to " Normal". . d. Train B slave relays will align for bypass breaker operation QUESTION: 049 (1.00) Which one of the following will result in illumination of the-"PCS CABINETS PWR SUPPLY FAILURE" annunciator for the Reactor Protection System (7300) Process Cabinets? a. Loss of the NYRS electrical bus, b. Loss of 24 VDC backup power to NSSS protection cabineti c. Energiz'.ng the Master Test Relay before placing the bistable test st ' itches in ' TES d. Loss af power to the Eagle 21 cabinet, l l- 1 j ! L T
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QUESTION: 050 (1.00) Which one of the following describes CONTROL ROOM indications that the Containment Iodine Reinoval System is in operation- following a LOCA inside containment? a. Containment spray pump suction pressure and discharge pressure at nominal operating values with both spray additive tank isolation valves ope b. Containment spray actuation signal present, at least one spray additive tank iso)stion valve open, and containment spray pump discharge flow rate at nominal operating valu c. Spray additive tank level decreasing, containment sump level , increasing, and containment spray actuation signal present, d. Cont.ainment pressure decreasing, containment humidity
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increasing, and containment sump level increasin , QUESTION: 051 (1.00) Which one of the following is the upper limit on hydrogen concentration in containment for starting a hydrogen recombiner? , a. 4%. %. %. d. 10%.
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__ REACTOR OPERATOR h _ h Page 31 QUESTION: 052 (1.00) Given the following conditions at Unit 1:
- A LOCA has occurred - Following the LOCA there was a fault on the Reserve Auxiliary Transformer (RAT) - "A" Emergency Diesel Generator (EDG) has started and is supplying its Emergency Bus - "A" EDG control air system piping has just ruptured Which one of the following describes the response of the "A" EDG to the loss of diesel control air pressure?
_ a. The engine will continue to run, but the output breaker will open causing the engine to trip on overspee b. The engine will stop and the output breaker will open deenergizing the Emergency Bu c. The engine will stop, _the output breaker will remain _ closed,_and the generator will motorize.- - d. The engine will continue to run, but al3 engine protective features will be inoperabl _ _ =----
.. REACTOR OPERATO (]) (]) .Page 32 QUESTION: 053 (1.00) Given the following conditions at Unit 1:
- Reactor Trip and Loss of Site Power has occurred - "A" Emergency Diesel Generator (EDG) starts, but output breaker fails to close - Bus 1AA02 is deenergized - Crankcase pressure indicates 3 psig - Lube oil pressure is 25 psig - Jacket water temperature is 200 degrees F: - Two minutes after starting, the "A" EDG shuts down Which one of the following caused the "A" EDG to shut down? -
a. Lube oil pressure, b. Crankcase pressure, c. Failure of output breaker to clos d. Jacket water temperatur QUESTION: .054 (1.00) An Emergency Diesel Generator is started for a test and is run for 48 hours at loads between 1500 KW and 3000 KW, Which one of the following describes the action required to be taken at the conclusion of the 48 hour test run? Deview all readings and trends; if satisfactory, shut down the engin b. If any cylinder exhaust temperatures have exceeded 900 degrees F, run at_3000 KW for an additional 2 hours,.and then'stop the engin c. Idle the engine unloaded for 2 hours or until temperatures (lube oil and cooling water) stabilize, and then stop the engine, d. Run the engine at 3500 KW load for 4 hours, and then shut down the engin . . . . . .
i > REACTOR OPERATOR () - (]) Page 33 i I l QUESTION: 055 (1.00) i Given the following conditions at Unit 1: ; r
- - Reactor plant in Mode 3 i
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- Containment temperature indications are: - TE 2563 (Containment Level 2) indicates 111 degrees F - TE 2612 (Containment Level C) indicates 117-degrees-F i - TE-2613 (Containment Level B) indicates 126 degrees F ,
Which one of the following is the temperature against which Technical- 3 Specification 3.6.1.5, Containment Air Temperature, must be applied? ; a. 111 degrees b. 117 degrees , c. 118 degrees d. 126 degrees r OUESTION: 056 (1.00) , Which one of the following lists the power supplies for the RHR. pumps? a. 1NA05 and 1NA0 b. 1NAA and 1 NA ' AB05 and 1BB07, AA02 and 1BA03, t
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QUESTION: 057 (1.00) . Which one of the following describes the response of the CCW system if # pumps, 1, 2, 5 and 6 are running when a Safety Injection is MANUALLY , initiated? ; a. All pumps will stop; pumps 1, 2, 3 and 4 will start at the appropriate sequencer ste : b. Pumps 1, 2, 5 and 6 will continue to run;-pumps 3 and 4 will t start at the appropriate sequencer ste c. Pumps 1, 2, 5 and 6 will continue to run, and no others will star d. Pumps 1 and 2 will continue to Imn, but pumps 5 and 6 will stop when pumps 3 and 4 sequence o ;
-QUESTION: 058 (1.00)
Given the following conditions at Unit 1:
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Reactor in Mode 4
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Cooldown to Mode 5 in progress Train A CCW Pumps 1 and 3 are running
, - - Train A CCW Pump 5 in standby ;
If a spurious CCW Surge Tank Lo-Lo level is detected by the. circuit that . controls CCW Pump 1, _w hich one-of-the'following describes the response .- of the Train A CCW Pump (s)? ; i a. CCW-Pumps 1 and 3-will continue to run until-a second lo-lo surge ta.;k level is actuated (a 2/3 coincidence) . b. CCW Pump i trips off; CCW Pump 3 continues to run, but CCW: Pump 5 will not autostart due to the lo-lo surge tank level signa c. CCW Pump 1 trips of f; CCW Pump 3 continues to run, and CCW Pump 5 autostarts_on low discharge header pressur d. CCW Pumps 1 and 3 trip off_and CCW Pump 5 is' prevented from autostarting due to the lo-lo-surge tank level signa : I
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REACTOR OPERt_ TOP O- O vase as QU29t dWr 059 (1<00) Which or.e of the following best describes the methods used to cool the Pressuriter Relief Tank (PRT) ? a. Recirculation through Reactor Coolant Drain Tank (RCDT) heat exchanger or spray from Reactor Makeup Water System (RMWST).
b. 'secirculation through RCDT heat exchanger or spray f rom Residual Hest Removal (RHR) tecirculatio c. Cpray frtm RMWST or venting to Waste Gas Decay Tan d. Spray f roc REF. Ce'drCulation or spray f rom excess letdow QUESTION: 060 (1.00) During recovery from a small break LOCA,_which one of the following-actions will rvduce the volume of a void formed-in the vessel head? -- a. Operate Pressurizur sprays, b. Operate Pressurizer heater c. Stop all Reactor Coolant Pump Stop Safety Injection Pumps.
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REACTOR OPERATOR' () (]) - Page 36
! - QUESTION: 061 (1.00)
The following plant conditions exist: ,
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- Unit 2 is in MODE Reactor water level is 22 feet above the vessel flang !
- RHR pump B is operating; RHR pump A is operable, but stoppe ,
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The following annunciators have just alarmed: ,
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CCW TRAIN A LO HDR PRESS annunciator
- CCW TRAIN A LO FLOW annunciator ,
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- CCW TRAIN A RHR PMP SEAL LOW FLOW annunciator
+ CCN TRAIN A RHR HX LO FLOW annunciator The Aux Building Operator reports flowing water, indicating a pipe rupture has occurred somewhere in the CCW pipin Which one of the following is a required action for these conditions?
a. Increase level in the Train B CCW surge tank to 65 percen , b. Reduce level in the Train A CCW surge tank to 35 percen c. Increase reactor water level to 23 feet above the vessel flange, d. Reduce reactor water level to 21 feet-above the vessel flang i QUESTION: 062 (1.00) Unit 1 is operating at 100% with Bank D rods at 218 steps. An ! electrical failure has deenergized the "C"'120V AC Vital Instrument Bus (ICY 1A) and now you note that the rods cannot be. withdrawn in auto or manual. Which one of the following is preventing outward rod motion? ; C-1, IR overpower rod sto . C-2, Power Range High Flux Rod Sto c. C-3, Overtemperature1 Delta-T Rod Sto ' d. C-4, Overpower Delta-T Rod Sto i 4 . ,, + ...w.,r,..wmm.,,4y,-..+,,,.,,,v- . g i,.v y ---.r , , , , .- , ~,,,_._..,__.,___-,-_mm,.Jugb_.-',,.-m ,M.e_,.-i-~._ ,,_ .-,
. _ . _ _ - _ . _ . _ _ _ _ _ - . - . _ _ .____._._._.__-._m . .
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REACTOR OPERATOR () (]) Page 37
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k QUESTION 063 (1.00)
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The heat flux hot channel factor limit calculation _in.nludes the term K(z), which is a normalized peaking factor that varies-as a function of core heigh Which one of the following is the reason-K(z) must be used in the heat flux hot channel factor limit calculation?
;
a. Compensates for the increased coolant temperatures that occur at higher core heights, b. Adjusts for the longer time delay in core reflood at higher core ,
'
heights following a LOC c. Adds a conservative uncertainty f actor since core flow becomes ! increasingly turbulent at higher core heights -
. .
d. Accounts for greater power production in the upper regions of the core near EOL due to axial flux shiftin , t
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QUESTION: 064 (1.00) One minute after a coincident reactor overpower accidcut/large-break LOCA occurred, containment pressure was noted to be steady at 12 peig and containment radiation was noted to be-steady at 2E+5 rad /h ; Therefore, trie use of Adverse Containment Conditions parameters was directed by the US Which one of the following sets of conditions would allow discontinuing the use of Adverse Contairiment Conditions? -
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CURRENT CURRENT CURRENT CONTAINMENT PRESSURE RADIATION LEVEL INTEGRATED-RADIATION (PSIG) (RAD /HR) LEVEL (RAD) .0 1.5E+5 3.5E+5 .5 1.5E+4 -1.~5E+5 .5 5E+4 1.5E+6 .0 SE+3 6.5E+4
,,
f y + . - n--- e-.* --r,-y,,- ..,,.,.-.,.e-m-. ,e. -..-i- _,,m-.y ww,.w.-n~n,e m- --.--+-1., 4
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REACTOR OPERATOR- l) _ () _Page.38 i i j QUESTION: 065 (1.00) ;
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An EOP step reads as follows: l Verify FW isolation !
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o MFIVs - SHUT o BFIVs - SHUT o MFRVs - SRUT t
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o BFRVs - SHUT The BULLETS ("o") indicate that: -[ a. the valves must be checked in the specified sequence (MFIVs; then DFIVs; then-MPRVs; then DFRVs). , t closing of these valves is more important than other actions without bullet c. the valves must all be checked, but the order in which they are checked is not importan .t these valves should already be in the correct position, assuming j that automatic actuations have occurred correctl > l
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QUESTION: 066 (1.00) WithLRCS pressure currently greater than 1375 psig, the following RNO is encountered while performing procedure 19010-C, " Loss of Reactor or Secondary Coolant,"
"
IF RCS pressure lowers to less than 1375 psig, THEN stop RCPs."
How.long is this conditional action directive applicable? a. Throughout 19010-C onl ; b. Only during this' step ' of 19010- c. Throughout the entire EOPs (FRGs and'ORGs), d. Throughout 19010-C_and after'any transition to another ORG.
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REACTOR OPERATOR O O vase n QUESTION: 067 (1.00) Which one of the following describes how REINITIATION of safety injection (SI) will occur if offsite power is lost following an SI - reset? a. SI will occur automatically upon the loss of offsite power, b. SI will occur only if manual action is taken by the operato c. SI will occur automatically when required SI actuation conditions exis d. SI can occur manually or automatically following the load shed - by the sequence QUESTION: 068 (1.00) In the post-accident monitoring system, the RVLIS hydraulic isolators are designed tot a. maintain tne containment pressure boundary in the event of a level sensing line brea b. adjust pressure in the level sensing lines to accommodate changes in containment pressure, c regulate flow to the transmitters to prevent damage from reactor pressure and level changes, d._ retain operability of RVLIS in the event of a loss of power _or instrument air.
/-
.. !
l
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r l REACTOR OPERATO O O vase 4o ! i QUESTION: 069 (1.00) One of the major directives contained in VEGP 19100, " Loss of All AC ! Power," is to: f r a. perform secondary depressurization in a rapid manner even if pressurizer level is lost during the depressurizatio b. refrain from resetting any SI signal to allow the SI sequencer to actuate when bus power is restore , c. enable the autostart of all ESP motor _ loads to ensure rapi recovery upon power restoratio i d. complete each step of the procedure prior to moving on to the next ste , QUESTION: 070 (1.00)- Following a Reactor trip on Unit 1, VEGP 19000-C, " Reactor Trip or , Safety Injection," is entered. When step 3.is reached, it is determined that 400 Vac busses _1AB04 and 19B16 are NOT energize Which one of th following describes the actions required at this point? a. Enter VEGP 19100;C, " Loss of All AC Power,".since one emergency bus on each train is deenergize b. Stop VEGP 19000-C at step 3 until the deenergized-busses are i reenergized from the diesel generato c. Continue with_VEGP 19000-C while attempting to restore power to the deenergized busses, d. Continue with VEGP 19000-C; no action is required for the deenergized busse .;
+
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_ . _ _ _ _ _ . _ REACTOR OPERATOR O O "ase 42 :
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QUESTION: 071 (1.00) Unit i has entered VECP 19100-C, " Loss of All AC Power," because NO AC - j emergency buses are energized. The following additional conditions exists
-
Steam driven AFW pump will not start ;
- All S/G levels are 10% (NR) j -
Reactor power is < 5% on all PR channels -t
" - Core exit TC's are 732 degrees F - Subcooling is O degrees F - RVLIS is not functional At this point the crew should ,
a. exit to VEGP 19221-C, " Response to Inadequate Core Cooling." -
,
b, exit to VEGP 19231-C, " Response to Loss of Secondary Heat Sink." , c. continue in VEGP 39100-C, " Loss of All AC Power."
d. implement VEGP 18038-1, " Operations from Remote Shutdown 1 Panels."
, QUESTION: 072 (1.00) With reactor; power at 7%, which one of the following sets of conditions requires a reactor trip on Unit 17 , a. PZR pressure 2285 psig; PZR level 18%; all SG' levels 42% NR; 1 Train "A" SI has occurred, PZR pressure 2335 psig; PZR level 25%; all SG levels 44%.NR; loss of 13.8Kv bus 1NAA has caused the loss of RCP's-1 and I c. PZR pressure 2035 psig; PZR level 94%; all SG' levels'40% NR; steam pressure negative rate bistables_are tripped on three steam line d._PZR pressure 1935 psig; PZR level 90%; all SG' levels 84% NR; Turbine Trip signal is present.
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REACTOR OPERATOR () () Page 42 i QUESTION: 073 (1.00) -i Given the following plant conditions:
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Pressurizer pressure 985 psig
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Pressurizer relief tank (PRT) pressure 5 psig ;
- PRT temperature 90 F . -
Reactor is shutdown i Assume ambient heat losses are negligible and the_ steam quality in the , pressurizer bubble is 100%. Also assume pressurizer and PRT-conditions do NOT chang Which one of the following PORV downstream temperatures would be caused by a leaking pressurizer PORV? , a. 230 F b. 260 F c. 300 F d. 340 F
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QUESTION: 074 (1.00) .
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One of the major actions in VEGP 19231-C, " Response to Loss of Secondary Heat Sink," is to establish reactor decay heat removal by bleed and feed operation Which one of the following describes the sequence of actions required to establish bleed and feed heat removal? " Bleed" is established by opening both pressurizer PORVs, and THEN " Feed" is established by initiating safety injection, b. " Feed" is established-by initiating safety-injection, and THEN
"81eed" is established by opening both pressurizer PORV c. " Bleed" is established by depressurizing a SG with its ARV, _and ,
THEN " Feed" is established by initiating AFW to_the.SG d. " Feed" is established by initiating AFW to a SG, Hand THEN
" Bleed" is established by depressurizing a SG with its AR .
-b
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- - . ~ . . , , . , , , ~ , . . . _ , . . _ , ._...w,,.._.._.._., , . . _ . . . . _ . . _ - . . . . . . - ~ _
_ _ _ ____._._ . _ _ _ . _ _ _ _ _ . _ . . _ _ _ _ _ _ . _ . . _ - - - . _ . . . - _ _ _ . _ . _ _ t REACTOR OPERATOR () () Page 43 I ' QUESTION: 075 (1.00) Why is a moderately small steamline break more likely to result in a I Pressurized Thernal Shock (PTS) condition than a large-steamline break? t - a. Cooldown is more rapid, t b. SGs will boil dry faste j, c. RCS pressure remains highe , d. SI injection flow is greate QUESTION: 076 (1.00) Refer to attached Critical Safety Function Status Tree (CSFST) F-0.4, Integrit Which one of the following describes the basis for limit A on Figure 17 . a. Gives the operator time to prevent a predsurized thermal shock conditio b. Indicates a potential for development of a flaw in the reactor pressure vessel if RCS temperature exceeds this limi ' c. Ensure-the "cooldown rate" of-the RCS is controlled to prevent
. permanent plastic deformation of the reactor pressure vesse d. Prevents'the growth of a flaw thaticould conservatively be present in the vessel wal .
A
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VEGP 19200- 9 7 og to
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Sheet 1 of 1
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INTEGRITY i l' i: G01019241-C~
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LE15 TMAM too*F IN 14 ust so ulwi YES amme G01019241-C E E
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m a - no N0 i TfWRATWE3 sREATER i - .r YES
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Peacemma no, envision n Pass s ~: VEGP of 10
- -
Sheet 1 of 1 3000 C T1 = 260 DEG. *F T2 = 290 DE *F P R ...................... E 2500
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S . U R E I 2000 N T1 T2 P S LIMIT A
1500 l 1000
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RED ORANGE YELLOW: GREEN 500 PATH PATH - - PATH PATH-
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GO TO l GO TO GO TO CSF 19241-C 19241-C 19242-C SAT
150 200 250 300 350 RCS WR COLD LEG TEMPERATURE IN (*F) !
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FIGURE 1 _ D
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l REACTOR OPERATOR () lll Page 44 QUESTION: 077 (1.00) Why are the RCPs required to be tripped at 1375 psig on a small-break LOCA? a. To eliminate RCP heat input into the RC L. To reduce inventory loss through the break, c. To prevent thermally shocking the reactor vesse d. To prevent RCP seal damage due to excessive seal injection flo QUESTION: 078 (1.00) After a large-break LOCA, appropriate actions of VEGP 19000-C, " Reactor Trip or Safety Injection," and VEGP 19010-C, " Loss of Reactor or Secondary Coolant," have been complete ECCS is operating in the Cold Leg Recirculation Mod RCS pressure is stable at about 200 psi How is core decay heat being removed? a. Condensation of reflux boiling in the steam generators (SGs).
b. Heat transfer between the RCS and the SGs due to natural circulation flow, c. Heat transfer between the RCS and the SGs due to forced - circulation flo d. Leakage of steam / water out the break and injection of cooled leakage wate __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - _ _ _ _ _ _ _ _ _ _
- REACTOR ~ OPERATOR- O O vaseT4s . QUESTION: 07 (1.00)
At sc$e time after-a large break LOCA, recirculation is transferred from-cold leg. recirculation to hot leg-recirculation.- The primary reason hot
' leg recirculation is used at this point rather than continuing _to use cold leg recirculation is:
a. to conserve RWST inventory, to prevent exceeding the boric acid solubility limit in the cor c. to stabilize RCS' temperatures when a pressurized thermal shock condition is imminen d. to collapse any voids that have formed in the steam geneEator tube QUESTION: 080 (1.00) A steam generator (SG) tube rupture has occurred, and VEGP 19030-C,
" Steam Generator Tube Rupture," is being executed. Before-the operator can commence the'RCS cooldown and depressurization per-VEGP 19030-C, the ruptured SG pressure must be checked-to be greater than'290 psig. Why is the 290 psig limit--imposed?
a. To ensure that the operator-can block the low st'eamline pressure SI signal, which would actuate below 290 psi _ b. To ensure RCS pressure will be'less than1 ruptured SG pressure after theLcooldown to stop primary-to-secondary _ leakage, c. To1 ensure a PTS condition for the reactor-vessel is not: developed during the'cooldown, d. To. preclude a return to criticality-during-the rapid RCS-cooldow , + _ _ kmmi iriu
. - - . _ . . . - - _ - . - -.. . - .- .- -.-~.-- ...-.~ .. . .
REACTOR. OPERATOR' h h Page.46: '
. QUESTION: 08 _(1. 0 0 )
The following plant; conditions exist:- _ Unit'l is at 25% powe RCP TRIP annunciator is-li .. .
- RCP LOOP 4 LOW FLOW ALERT annunciator'is li '
- RCP 4 MTR OVERLOAD annunciator is ~ 11 Which one of the following actions is an IMMEDIATE ACTION for this event I per VEGP 18005-C, " Partial Loss of-Flow"?
~
a. Check RCP 4 has tripped and attempt to restar t b. Close Loop 4 spray valve (PIC-0455B).
c. Verify ~ pressurizer-level is trending to progra d. Verify No. 4 steam generator level is trending-to 65%. QUESTION: 082 (1.00) , Which one of the following conditions requires emergency'bora .on to be performed? a. Reactor Engineering reports that shutdown margin-equals 1.3% delta-k/k while in MODE 1.
> b. Intermediate range power (amps) has doubled in three minutes while in MODE Control rodlbank D stepping out without operator action wh'ile in MODE d. SOURCE RANGE HIGH FLUX AT-SHUT DOWN annunciator in: alarm while in MODE 3.
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- REACTOR OPERATOR O & rase 47
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QUESTION: 083. -(1.00) VEGP 19030-C, " Steam Generator Tube Rupture," once entered, should he performed'in a timely manner tot-a. prevent water from entering the main steam line b. prevent overpressurization of the ruptured steam generato c. prevent formation of a vapor bubble in the reactor vessel hea , d. prevent having to stop the RCPs prior to initiating-the maximum rate cooldow QUESTION: 084 (1.00) With Unit 1 in MODE 1, the following annunciators have just alarmed:
- ACCW RCP 3 CLR LO FLOW annunciator - ACCW RCP 3 CLR OUTLET HI TEMP annunciator - ACCW RTN HDR FROM RCP LO FLOW annunciator According to VEGP 18022-C, " Loss of Auxiliary __ Component-Cooling Water,"
which one of the following conditions requires tripping No. 3'RCP?
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a. Seal water outlet temperature at 200 degrees I b. Motor stator winding temperature at 200 degrees c. Motor lower radial bearing temperature at-200 degrees Pump lower seal water bearing temperature at 200 degrees F.
x _ . .
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- - _ _ _ - - - - - _ 'REACTORi-OPERATOR h Page:48'
QUESTION: 085. '_(1. 0 0 )
.The following-plant conditions exist: - Unit 1 is:at 45 percent-powe TURB CNDSR LO VAC annunciator is li Condenser vacuum gauge reads 19.5 inches H Which one of the following should have occurred?
a. Turbine runback, b. Turbine trip onl _ c. Reactor trip causing a turbine tri d. Turbine trip causing a reactor tri . QUESTION: 086 (1.00) The following plant conditions exist:
- The diesel fuel oil storage tank has ruptured._ _ . - A large' diesel fuel oil-fire is. burning near the storage tank'. - Smoke has entered the intakes for the control room (CR) - ventilation syste ~
Which one of the following describes how the-CR HVAC is reconfigured to _ protect the shift personnel from the smoke hazard? _ a. Operators manually shift CR HVAC to the purge mode, b. Operators manually shift CR'HVAC to the isolation mode, c. Smoke detectors cause automatic shift of CR HVAC toithe purge
-
mode, d. Smoke detectors cause automatic shift of CR HVAC to the-isolation mod l
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- - - . . - . . - ~ . . . - - - - . - . - - - - -- .. - - .~.. - - - . . - . - . . ---
REACTOR OPERATOR (f () Page 49
. QUESTION: 087 (1.00) -
The following plant conditions exist:
- Unit 1 is at 100 percent power with pressurizer. pressure control in Aut Pressurizer pressure channel selector is in the 455/458 positio Pressurizer pressure transmitter PT-458 fails tol2390 psi ,
Which one of the following actions or conditions will STOP further plant-degradation from this-event? a, Pressure transmitter PT-456 senses less than 2150 psi Pressure transmitters PT-455 and PT-457 sense less than--2185-psi c. The operator manually closes PORV PCV-455 d. The operator manually closes spray valves PCV-455B and PCV-455 . QUESTION: 088 (1.00) If rod control is in AbrO when symptoms of a possible narrow range temperature instrument tailure-are noted, what action should be taken by the operator BEFORE the control rods are placed in MANUAL? a. Verify the rods are stepping I b. Verify the temperature instrument is-NOT in tes c. Verify turbine runback or load shed is NOT in_ progress, d. Verify reactor trip should NOT have occurred-due to the failure.
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D Page.;50
, - REACTORLOPERATOR h-QUESTION: 089 - (1. 0 0 ) - -
1The_following_ plant-conditions exist:
- Unit 2Lis at 100-percent powe Letdown orifice valve HV8149C.is ope Pressurizer level is increasin T-avg is constan LTDN HX OUT HI PRESS annunciator li Charging flow is decreasin All control systems are in automati Which one of the_following a ents would cause these conditions?' -
a. Orifice-isolation valve HV-8149C. closin b. Orifice isolation valve.HV-8149A openin c. Letdown isolation valve LV-459' closin d. Letdown pressure control valve PCV-131 closing. . s QUESTION: 090 (1.00) _The following conditions exist:
- The unit is in Mode RHR Train A is in operatio Reactor vessel level is at 188 feet (mid- loop)' .
-RHR Heat. Exchanger Bypass Flow Controller (FIC-0618A) has_just: failed,;
. causing RHR flow to increase'to maximum. If?no operator action is taken
.
'and-RHR flow remains at-maximum, which one of1the followingLwould occur to.cause a loss-of RHR?
'
a. Pump _overspeed_ trip from runout due_to low discharge-pressure.- b. Pump overcurrent trip due_to the high discharge-pressur . c. Loss of pump suction due to gasLentrainment in the loop'suctioni lin d. Lossiof pump suction due to low net positive suction head.
. 4g- g w W - - t s-1- = r 4 - M
REACTOR '_O' ERATOR : 0; LO1 Fase s1;
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: QUESTION: 091 (1.00)
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Which one of_the following describes the final Tavg-change'as_a result of a turbine runback .(rods in auto) versus the- final Tavg1 change as a'- result of an uncontrolled continuous rod insertio a. Final Tavg for'both will INCREAS Final Tavg for both will DECREAS a c. Final Tavg will INCREASE for a-runback while final-Tavg will-- DECREASE for an uncontrolled-continuous rod insertio Final Tavg will DECREASE for a runbackLwhile final Tavg will INCREASE for an uncontrolled continuous rod insertio ' QUESTION: 092 (1.00) Which one of the following RCS leak rates is within allowable limits per
* -the plant Technical Specifications?
a. 8_gpm identified leakage, b. 8 gpm unidentified leakage, gpm through RHR suction valve (HV-8701A). gpm steam generator tube leakag . _
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4
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3 .g- - -- y 9 w---my- y g y yy' y re ya -ry yr --r"
, .. - _ - - ~ . - .. . ~ . - - . _- - .. - . - ~.- - .- . ' REACTOR OPERATOR h Pag'e-52 QUESTION: 093' .(1.00)
Control. rod H8 has just been withdrawn 15 steps to realign it with control bank (CB) D in accordance with VEGP 18003-C, " Rod Control System- . Malfunction." After the rod was realigned, the P/A converter for CB-D-was NOT reset to the original bank heigh ~ . Which one of the following will occur because of this oversight?
- ROD CONTROL NON URGENT FAILURE alarm when CB D is inserted, b. ROD CONTROL URGENT FAILURE alarm when CB D is withdraw c. Spurious ROD DEV/ RADIAL TILT alarm Improper ROD BANK.LO LIMIT alarm QUESTION: 094 (1.00)
Which one of the following explains why it is important to shift rod control to. AUTO during a loss of main feed pump with plant power at 90%? a. To reduce Tavg to 557 degrees F, thereby increasing the DNB b. To ensure Tavg is not reduced below the minimum temperature for criticality, c. To maintain Tavg accurately on program for the indicated steam flow, d. To match Tavg to Tref more quickly than manual rod 1 control ca .. . - - , -.
.. . -~ . - . -. . - - - . . . . - . - - - .. . . . ..
REACTOR: OPERATOR -Page 53 QUESTION: 095 (1.00) Given the following.' conditions on Unit.1:
-- ' Unit 1 is at 100% powe Rod control is in manua SERVICE AIR HDR LO PRESS annunciator has just alarme Instrument-air (IA) pressure is fallin Which one of the following describes the action to be taken? If IA pressure drops below 100 psig, isolate service air from IA.
. If IA pressure drops below 80 psig, reduce reactor power toiless than 50%. If IA pressure drops below 80 psig, verify automatic isolation of turbine building IA, If IA pressure drops below 70 psig, trip _the reacto QUESTION: 096 (1.00) Which one of_the following describes No. 1 steam generator (SG-1) response to a trip of No. 1 reactor coolant pump at 25% reactor power? a. SG 1 level will " shrink" and then increase due-to FRV leakage.- b. SG 1 will continue to steam at 90% of the pre-event _ rat c. SG 1 level will decrease to the.lowLlevel trip setpoint,
- SG 1 pressure will increase to' saturation pressure for RCS Tavg.
't ! y m-
' REACTOR: OPERATOR () () Page 54' iQUESTION: 097 (1,00)
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The following pl.nt-conditions exist:
- Unit 1 has tripped from 100% due to a-loss of off-site powe Both diesels have started and the sequencers are loading-the buse seconds after sequencer start, SI is inadvertently actuate Which one of the following explains the subsequent response'of the Unit 1 sequencers to-the SI actuation signal?
a. The sequencers taust be reset manually before the SI loads will' begin sequencin b. The sequencers will reset immediately and start sequencing the SI loads, c. The sequencers will reset in 20 seconds and start sequencing the SI load . After 20 seconds the sequencer.a will start sequencing SI loads not' started previously by the UV sequence QUESTION: 098 (1.00) The following plant conditions exist:
- Unit 1 is in MODE Fuel Handling Building Ventilation Isolation has actuate Which one of the following signals could have caused this. event?
a. LOW air flow in the nornal Fuel Handling Building HVA b. LOW air flow in the. fuel pool area recirculation unit .c. Fuel Handling Building pressure exceeds. atmospheric pressure, d.-Spent Fuel _ Pool temperature reaches 132Edegrees l
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REACTOR OPERATOR O C ve9e ss 1 QUESTION: _099- (1.00) With the plant at :100% power, which .one of the Ifollowing can result from an extended loss of utility water if no corrective action is taken? a. 50% loss of condensate storage tank inventory, t b. Loss of MFPs on high lube oil temperatur c. Main turbine trip on low condenser vacuu d. Low water level alarm on boronometer shield tan QUESTION: 100 (1.00) A small fire causes the loss of bus 1AA02. The'controlLroom operators-should perform which one of the following IMMEDIATE operator actions:- a. Emergency stop A diesel generato b. Commence a plant shutdown-to hot standb c. Check A diesel generator running and attempt to energize 1AA02 after the fire is out, d. Attempt to close A diesel generator output breaker; if it won't close, normal-stop A diesel generator.
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> (********** END OF EXAMINATION **********)
J p ,y .,y - - . ------- -
- - - __ _ . _ _ _ _ - _ - _ _ - _ _ - - _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ - _ _ _ _ , REACTOR OPERATO .Page 56-ANSWER: 001. (1.00) REFERENCE:
.LO-LP-63008-07-C, pg. 7 KA 194001K105 [3 .1/ 3. 4 ]
194001K105 ..(KA's) _ ANSWER: 002 (1.00) REFERENCE: VEGP 00304-C, pg. 10 KA 194001K102 - [ 3 . 7 /4 .1] 194001K102 ..(KA's) ANSWER: 003 (1.00) REFERENCE: VEGP 00304-C, pg. 22 KA 194001K102 [3.7/4.1] 194001K102 ..(KA's)
- - . . _ . . _ . - - - --
.
REACTORLOPERATOR- O O : veee 52 ( S ANSWER:. 004 (1.00)
^ REFERENCEi LO-LP-63308-09-C, pg. 10 KA-194001K101 [3.6/3.7)
194001K101 ..(KA's) v ANSWER: 005 (1.00) ~ REFERENCE: ._ VEGP 00404-C, pg. KA 194001A103 [2.5/3.4) 194001A103' ..(KA's) ANSWER: 006 (1.00) - '
- - ) REFERENCE: .VEGP 10000-C, pg.-15 KA-194001A1111 [2. 8 /4.1) .
194001A111 ..(KA's) esh I
- -
_..-__ _ - __. . _ - _ - _ - . _ _
- REACTORIOPERATOR'. h h Page'58
,
ANSWER: 007 ( 00 ) f _ ~
, REFERENCE:
VEGP 10004-C, pg. 3 _ KA 194001A103- [2.5/3.4] 194001A103 ..(KA's) ANSWER: 008 (1.00) REFERENCE: VEGP 10019-C, p '
- -l KA 194001K101 [3.6/3.7) -
j 194001K101 ..(KA's) -ANSWER: 009 (1.00) REFERENCE: VEGP 00920 ", pg. 7 KA 194001K103 (2.8/3.4] 194001K103 ..(KA's)
.
., .. - .. . - . . . . . . ~ . .-.n.... _ . . . . . . - . . . . . .-- . - . . ~
JREACTORIOPERATOR- -
-
Page-59
'
ANSWER: -010- (1.00) d ,-
'
REFERENCE:
~
VEGP 00920-C, pg. 3 KA 194001K104 [3.3/3.5)-
,
194001K104 ..(KA's) ANSWER: 011 (1.00) . REFERENCE:
-VEGP-00308-C, Sect. 2.3, 4.1. KA 194001K104 [3.3/3.5)
194001K104 ..(KA's)
.
ANSWER: 012 (1.00) REFERENCE:
.VEGP 00920-C, pg. 7-KA-194001K103 [2.8/3.4]
~
194001K103 ..(KA's)- (
>
-
.,
, . - - , . -. . . - . . . - .- .-. . . . . .
REACTOR;OPERATORL
] h - Page 60:
.
,
: ANSWER: - 013-.(1.00)
_ _
,
REFERENCE:
-VEGP 00920-C, pg.:5 d KA-194001K103 (2. 8/3. 4)
194001K103 ..(KA's) ANSWER: 014 (1.00) . REFERENCE:
*
- LO-LP-27101-16-C, pg. 21, L.O. 9 KA'001000A102
'
[ 3 .1/ 3 . 4 )
001000A102- ..(KA's).
. ANSWER: 015- (1.00)
, ' REFERENCE:
LO-LP-27101-16-C,~p. 26 KA 001000K402 [3. 8/3. 8)
=001000K402 ..(KA's)-
.
%
g e e a u , - --.--.,,.ca. # -- .2
- --
REACTOR _OPERATORT .' O P se 51
$ ANSWER: 016: (1.00) . _
REFERENCE: LO-LP-27102-10, pg.-13, KA 001000A106 (4 .1/4 . 4 ] 001000A106 ..(KA's).
C ANSWER: 017 (1.00)
, REFERENCE: -LO-LP-27012-10, pg. 16, L.O. 9 KA'001000K504 (4.3/4.7)
001000K504 ..(KA's) ANSWER: 018 (1.00) ' - . REFERENCE: < LO-LP-16401-14, pg. 12 JKA 003000K201 [3.1/3.1) 003000K201 ..(KA's)
. . . . .
._ . , . . . _ _ . - - _- - _ . - _ _ . _ , . - . . _ - . - . . . _ _ - _ . . _ . - _ . .
'
; REACTOR-OPERATOR- =
. Page;62-i ANSWER: -019 .: (1. 0 0 )- -
_
'
C.- REFERENCE:,- LO-LP-16401-14, pg. 19 KA 003000G015 [3. 8/4. 0)
-s 1003000G015 ..(KA's) ..
ANSWER: 020 (1.00) REFERENCE: LO-LP-16401-14, pg. 11, KA 003000K201 [3.1/3.1) 003000K2Gl'- ..(KA's) ANSWER: 021 (1.00)
- REFERENCE:
LO-LP-09101-08, , L.O. 8-KA 004010A204 [3.6/4.2] . 004010A204 ..(KA's)-
.
T '
-, .-. . _ _ . _, , _ , . _ . - . . _ _ _ , _ . - _ _ _ , _ , . - _ , . . . . , _ . _ . , , . , . . _ _ _ _
_ . . . . . . . . . _- .. . - . _ . . _ . .. REACTOR l OPERATOR- h Page'63' _
-
_
' ANSWER:- -.022 .(1.00) ' .a.- - --REFERENCE:
VEGP 13011-1, pg. 3 'l KA 006000K603 (3.6/3.9) 006000K603 ..(KA's) ANSWER: 023 (1.00) d REFERENCE: LO-LP-09501-05,.pg. 11, KA 004000K104 (3.4/3.8] 004000K104 ..(KA's) . -ANSWER: 024 (1.00)
. REFERENCE:
VEGP 13009, Section KA.-004000A207 -(3.8/3.9] 004000A207 . . .( KA ' s ) u l l l ..
, , , . . _ . . . , . . . . _ .. _ , .;..,-. . ., c
iREACTOR! OPERATOR.- 0; O vase 64 ANSWER: - 025- (1. 0 0 )-
-; d . -
REFERENCE:
'LO-LP-28103-13-C, pg. 32, KA 013000K103 [3.8/4.1)
013000K103 ..(KA's) ANSWER: G .S. C (1.00) . REFERENCE: LO-LP-28101-10-C, pg. 13, KA 013000A301 [3.7/3.9) 013000A301 ..(KA's) ANSWER: 027 (1.00) REFERENCE: LO-LP-36102-04-C, pg. 14, L.O. 7-KA 017020A201 -[3 .1/ 3 . 5 ] 017020A201 ..(KA's)
- - --
. __ _ _ - _- - _ ___ _ - __._-__ _ _ _ _ _ _ _ _ _ - - - .. . ' REACTOR OPERATOR'^ .
h Page-65 LANSWER: 028: (1.00) . .
. REFERENCE:
i-LO-LP 29130-04,. pg. 6, L.O. 2 KA 022000A301 [4 .1/4. 3 ] 022000A301 ..(KA's) _ ANSWER: 029 (1.00) REFERENCE:
'
LO-LP-18201-13-C, pg. 13, L.O. 7 KA 059000K419 (3.2/3.4] 059000K419 ..(KA's) _ ANSWER: 030- (1.00) -
. '
REFERENCE: LO-.LP-20101-17-C, pg. 11, L.O. 3 KA.061000K404 . { 3 .1/ 3 . 4 ] 061000K404 ..(KA's) l N
. .
- .- . . . .- ., - . . ~ . . ~ . - _ , - . . . . . . . - . - . - ~ - . . . , . . .
. . . - - . .
REACTOR' OPERATOR. - = h _Page'66;
'
-
i
'
ANSWER:- 031: (1.00).
- REFERENCE:
/LO-LP-20101-17-C,~pg. 12, KA'061000K103 [3.5/3.9]-
061000K103 ..(KA's)
.
ANSWER: 032 (1.00) REFERENCE: LO-LP-20101-17-C, pg. 12, L.O. 4 KA 061000A102' [3.3/3.6] 061000A102 ..(KA's). .
!
!
ANSWER: 033 (1.00) REFERENCE: LO-LP-O'1001-04,fpg. 11, O. '6
=.KA~062000A401 ~(3.3/3.1]
062000A401' ..(KA's) , { r
r - ..,N ., - , . . . . . , . , ,-
.. ,. . - . . . . . . . . . - ... . . .. . . - . . - . .
-
- . . .
. .
. REACTOR OPERATOR-G O > se 57 -
.
ANSWER: 034 , (1.00) :
-
REFERENCE: LO-LP-32101-15-C, pg. 25, L.O. 11
,
KA 072000G008 (3.1/3.1) 072000G008 ..(KA's) ANSWER: 035 (1.00) , REFERENCE: LO-LP-16303-12-C, pg. 9, , L.O. 5
; KA_010000K403 [3.8/4.1] -010000K403 ..(KA's) ' ANSWER: 036 (1.00) REFERENCE: .LO-LP-16303-12-C, p , L.O. 5 KA 010000A403 . [4. 0/3. 8']
01c000A403' ..(KA's) , e
} . . . - - - - a - .. , . , y nt-
.. . . .~ . _ _ - . - _.- ._ __ _ .--_ _ _ _ - . _ - . __ . _ . . _ _ . __ _
,
REACTOR-IOPERATOR :;h' -Page'68
:9
-
,
c')
' ANSWER: 037 ~ - (1; 00)-
[
a, r
!
REFERENCE: LO-LP-16302-08-C, pg. 15, L.O. 2 KA 011000A211 [3.4/3.6] _ 011000A211 ..(KA's)
, - .
ANSWER: 038 (1.00) , '
.? -REFERENCE:
- LO-LP-13001-11-C, p , L.O. 2 KA 006000G004 [3.5/3.8) ,
:3 006000G004 ..(KA's) >
',
ANSWER: 039 (1.00)
-;
a,
.
REFERENCE': LO-LP-16001'11, pg. 23-24,.L.O. 1
~ KA002000K106 - [3.7/4.0]
002000K106 ..(FA's, l
i y ,p - y .,_y,
- =.+,e + . , . -. ,,
__ _. .. . - . _ . . _ . . . _ . . . . _ . _ . . . . . . . . ._ . 'I
.. REACTOR' OPERATOR- '_ , Page 69I
-
1 ANSWER:- 040 (1.00) - s 4,
, REFERENCE:
i ' LO-LP-16701-08-C, L.O. 5 KA 002000K402 - {3. 5/3. 8)
,
002000K402 ..(KA's)
,
ANSWER: 041 (1.00) ' REFERENCE: LO-LP-01201-11-C, p , L;O. 3 KA'063000K301 [3 . 7/4.1) 063000K301 ..(KA's)
.
. ANSWER: 042 (1.00) REFERENCE: ,
,
LO-LP-25702, pg. 20; VEGP,ARP 17005 KA'033000K303 (3.0/3.3]
. .
033000K303 ..(KA's) _.
..
$
w - 4 ,+ , - 1
, _ . . , , _ . - . - .- - - - - .---..~.- - - . - . - - . . . . - . _ _ . . . - . . - . _ - .
!
REACTOR OPERATOR h Page 70
?
>
ANSWER: 043 (1.00)- , !
!
;
REFERENCE: ! t LO-LP-25201-15 C, pg. 21, l r KA 034000A201 (3.6/4,4] f 034000A201 ..(KA's) I i
- ANSWER: 044 (1.00)
t t
. - REFERENCE: - I LO-LP-17101 07-C, pg. 14, KA 015000K301 (3.9/4.3]
t 015000K301 ..(Ya's) ANSWER: 045 (1.00) REFERENCE: LO LP 27201-08, pg. 11, L.O. 6 KA-014000G015 (3.3/3.5) , 014000G015 ..(KA's) l
,
!
'
.
E I l- ,
, :.' ~
t
...,m..-,-.., . . , _ _ _ . - . . , _ _ . _ , _ . - , - , __,...._..._,,._._._.-,,..._..,,-,,,_.--.._..,-,.:,._,-.._.-....._,1_
-
. .. _
-
.. - REACTOR OPERATOR (]) () Page 71-ANSWER: 046 (1.00) - .
REFERENCE: LO-LP-17201-08-C, pg. 11, KA 015000A303 [3.9/3.9) 015000A303 . . pa ' s )
- - ANSWER: 047 (1.00) REFERENCE:
LO-LP-28103-13-C, pg. 23, KA 013000A402 (4.3/4.4] 013000A402 ..(KA's) ANSWER: 048 (1.00) REFERENCE: LO-LP-28101-10 C, pg. 12, KA 012000K401 [3.7/4.01 012000K401 ..(KA's) \
:l
. __._- .,__ _ _ _ _ _.._. . _ ._.__..--_.._-_._ ~ _ _ _ _ . . . . _ . ~ . - . . _ . . . _
REACTOR OPERATOR O O vase 72 i
.
.I ANSWER: 049 (1.00)
]
b.
. REFERENCE: : LO-LP 28102 08, pg. 8, L.O. 2 I i Va 012000G008 [3.9/3.8) !
'
012000G008 ..(KA's) r
<
ANSWER: 050 (1.00) REFERENCE: LO-LP-15101-10, pg. 15 KA 027000A401 [3.3/3.3)
!
. 027000A401 ..(KA's) 'f . ANSWER: 051 (1.00) ! REFERENCE:
;
LO-LP-29140-07, p , L.O. 2-KA 028000G005 [3.0/3.6)-
.
'028000G005
'
"
..(KA's)
>
%
,
.
.,
. E
.-,
t
' Y - , e. - + . . . , . . - . . . , .- . . - - . , , . , --. - . _, -. .4.,'. . . , . , - .< .- - , , ,~,s , , , , e . w -., en ~....
. . . . . . REACTOR OPERATOR O O vase 73 ANSWER: 052 (1.00) d, REFERENCE: LO-LP 11201-14-C, pg. 55, LO 15 KA 064000G007 (3.4/3.6) 064000G007 ..(KA's) _ ANSWER: 053 (1.00) REFERENCE: LO-LP-11104-0-C, pg. 17, LO 9 KA 064000A101 (3.0/3.1) 064000A101 ..(KA's) ANSWER: 054 (1.00) . REFERENCE: LO-LP-11202-15-C, pg. 60, L.O. 13 KA'064000A406 (3.9/3.9) 064000A406 ..(KA's)-
. . - . . :
. . . . . . REACTOR OPERATOR () () Page 74 ANSWER: 055 (1.00) REFERENCE: LO LP-29120-04-C, p , KA 103000A101 (3.7/4.1) 103000A101 ..(KA's) _ ANSWER:- 056 (1.00) REFERENCE: LO-LP-12101-30-C, pg. 11, L.O. 5 KA 005000K201 [3.0/3.2) 005000K201 ..(KA's) ANSWER: 057 (1.00) ~
- REFERENCE:
LO-LP-10101 11, pg. 16, \ 008030A304 (3.C/3.7) 008030A304 ,.(KA's)
. . . .
.- .- . . - . . . . - - . - ~ . . ~ . ~ - . ~ . . . . - . . - . - . - . . - .-..- ... . . ..
!
REACTOR OPERATOR O O rese 25' 1
!
,
MfSWER: 058 (1.00) t '
,
'
!
REFERENCE:
;
LO-LP-10101-11, pg. IS, KA 008000A301 [3.2/3.0) ,
:
000000A301 ..(KA's) i ANSWER: 059 (1.00) REFERENCE: , L LO-LP-16301-12 C, p. 24, 25, L.O. 13 ' KA 007000K401 [2.6/2.9) ' 007000K401 ..(KA's)
.
ANSWER: 060 (1.00) . REFERENCE: > Steam Tables KA'010000G013 [3.5/3.7)- ' 010000G013 ..(KA's)
.
__ - h ANSWER: 061 (1.00)
, ,
>
gn,+,y, .,--, ,7 e r., ry r, - -v-,- , --
<-,,,,n<n-. .-e ,----,,----.w--,,,,, - - - . - ,-.-n- - . , - -- , . , , ,- -----.--,---,,-~~-r--
_ ~ . . . . . . . - . . . _ - . - - . . . - . - . . . - . . . . . .. -. - --. . . - - - . . - - - . ~ ~ . . _ . . . REACTOR OPERATOR O O vase 2a REFERENCE: ' VEGP Tech Spec 3.9. I VEGP.18020-C, rev. 5, pg. 2 KA 008000K301 (3.4/3.5) ' 008000K301 ..(KA's) ;
,
T ANSWER: 062 (1.00) ; .l
'
REFERENCE: LO-LP-60324-01-02,-pg. 20 KA 000057A217 [3.1/3.4) 000057A217 ..(KA's) ANSWER: 063 (1.00) . REFERENCE: Ops Trng MCL CL-36 1.06 KA 000001G003 -{3.3/3.9) 000001G003 ..(KA's) ANSWER: 064 (1.00)
.
, t v y - 41,v m s e g+,e -'w, ,- w ww - v - - ivv-w *-mE+rw-w-wmy---+-+- w ww-e ev rwve - e sw, e-v =~ o w--e,-r--We"-
. . - . . . . -. - - . - .. - - - - - - - - _ .. - . - - - .. _ - _ .
REACTOR OPERATOR-0 0 vase >> ; i
.
REFERENCE:
;
LO-LP-37002-09-C, pg. 6,7 l
KA 000011G010 [4.5/4.5) i 000011G010 ..(KA's) . t ANSWER: 065 (1.00) r ,
!
REFERENCE: , LO-LP-37002-09-C, pg. 6 KA 000069G012 (3.5/3.5) 000069G012 ..--(KA's).
ANSWER: 066 (1.00)
, REFERENCE: -LO-LP-37002-09 C, pg. 6 KA 000009G012 [4 . 0/4 .1]
000009GO:12 ..(KA's) ANSWER: 067 (1.00) . . . _ _ . . _ , , _ . . . . . - - _ . . _ . _ . _ _ _ _ _ . . . . , _ . _ _ , . , _ . _ _ . _ . _ _ . - _ . _ _ . ~ . - _ _ . _ _ _ ....__.~.. .,_.. . .-
_ . . . . . REACTOR OPERATOR (]) () Page 78 REFERENCE:
-VEGP 19011-C, pg. 2 KA 000056A203 (3.8/3.9)
000056A203 ..(KA's) ANSWER: 068 (1.00) REFERENCE: LO-LP-37003-07-C, pg. 12
}UL 000074A208 (2.5/2.5) -000074K208 ..(KA's)
ANSWER: 069 (1.00) REFERENCE: LO-LP-37031-09-C, pg. 7, VEGP 19100-C, pg. 19 KA 000055K302 [4.3/4.6) 000055K302 ..(KA's) ANSWER: 070 (1.00) .
_ _ _ _ _ _ _ _ _ _ _ _ . _ . _ . . _ . _ _ . _ _ -
. _ _ - -
i REACTOR-OPERATOR O O rase 79-i
!
REFERENCE: , VEGP 19000+C, pg. 2 , KA 000007A202- (4.3/4.6) , 000007A202 ..(KA's) .{
!
ANSWER: 071 (1.00) l i l l REFERENCE: ;
;
LO-LP-37031-09-C, pg. 6 ;
,
KA 000055G012 (3.9/4.0) 000055G012 ..(KA's) ~ ANSWER: 072 (1.00) REFERENCE: VEGP 19000-C, Att. A KA;000029A209 (4.4/4.5) 000029A209- ..(KA's) ANSWER: 073 (1.00) ~ _:_.-_.-.-..,...-..-.-
__ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ . REACTOR OPERATOR O O ra9e 80 REFERENCE: Steam Tables KA 000008K302 [3.6/4.1) 000008K302 ..(FA's) ANSWER: 074 (1.00) REFERENCE: LO-LP-37051-10-C, pg. 9 YA 000054K305 [4.6/4.7) 000054K305 ..(YA's) ANSWER: 075 (1.00) REFERENCE: Ops Trng MCL CL-37 3.47 KA 000040K104 [4.1/4.4] 000040K104 ..(KA's) ANSWER: 076 (1.00) d.
_ _ _ _ _ _ - _ _ - - - _ _ _ _ _ - - _ _ _ - _ - _ _ _ _ _ _ _ - _ _ _ _ _ _ _ - - _ _ _ - _ - _ _ _ - - _ _ _ _ _ _ _ _ _ ______ -
. _ _ _ . _ . _ _______._-_-_.____...._..._.___-_.._._._.._.m._._._____, - - - .
REACTOR OPERATOR O O rese 81 REFERENCE: LO-LP-37071-03, pg. 8 KA 000040K101 [4.1/4.4] 000040K101 ..(KA's) ANSWER: 077 (1.00) -I i l l REFERENCE: LO-LP-37111-09-C, pg. 10 I KA 000009K323 (4.2/4.3] 000009K323 ..(KA's) , F ANSWER: 078 (1.00) d.- ,
,
REFERENCE:
,
VEGP 19010-C KA 000011K305 (4.0/4.1]
'000011K305 ..(KA's) ~ ANSWER: 079 (1.00). , 'i
! ,
'h
,_a.._..._,.--..a.-._,,-~:,,-
-
, . , , _ . _ , , ....-..-,,_....,_--..- . _ , . . . , . - _ , _ . . _ . . ..--._.._. 4 .. : ._-- - - - - . ~
. . . _ _ . _ - _ - . . .. _ ..- -.. .-_---.._.- _ .. .. .. . -.. . . - .,- . - - _-
'
REACTOR OPERATOR O O rese 82 REFERENCE: , LO-LP-37114-08, pg. 6 i RA 000011K313 (3.8/4.2] ; 000011K313 ..(KA's) ANSWER: 080 (1.00) t , REFERENCE: Ops Trng MCL CL-37 4_.17 KA 000038K306 [4.2/4.5) 000038K306 ..(KA's) 5 ANSWER: 081 (1.00) REFERENCE:
,
VEGP 18005-C, pg. 2
'
KA 000015G010 [3.4/3.4) 000015G010 ..(KA's)
.
ANSWER: 082 (1.00)
..
. .,
._
a
i
;
a . -- , . . - - . .-......,-.--.__.:-.-.-,...-,;.-a.-~.~ ,...--._4 - . A .-
REACTOR OPERATOR O O rese 82 REFERENCE: VEGP 17010-1, pg. 19 KA 000024K301 [4.1/4.4) 000024K301 ..(KA's) ANSWER: 083 (1.00) REFERENCE: VEGP 19030-C, pg. 14 KA 000038A144 (3.4/3.4) 000038A144 ..(KA's) ANSWER: 084 (1.00) REFERENCE: VEGP 18022-C, pg. 4-KA.000026K303 (4.0/4.2) 000026K303 ..(KA's) ANSWER: 085 (1.00) i
'
- , ~ , , ,e. .w....-,-. __.__-__.___4,..-..,,,..
- -..__.e-,-.-...,_m,- . . . . . _ . .
_ . , _ _ _ . _ . _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ . . . . - - . _ _ _ _ . _ . . _ . . . _ _ . _ _ _ . . _ . . ~ _ _.- . . _ _ _ . . _ . _ _ . _ . _ . . . . _ . ._ REACTOR OPERATOR O O vase 84 l: REFERENCE:- VEGP 17019-1, pg. 11
[3.9/4.1)
'
KA 000051A202 t
'
000051A202 ..(KA's)
;
,
i , ANSWER: 086 (1.00) i t : REFERENCE:
:
VEGP LO-LP-23101-10-C, pg. 12 l KA 000067A105 [3.0/3.1] =- i 000067A105 ..(KA's) t
.
ANSWER: 087 (1.00) REFERENCE:
^
VEGP Logics, Fig. 7.2.1-1, shts. 11,-18, & 19 VEGP LO LP-60301-07-C, pg. 14 > KA 010000K403 (3.8/4.1)
,
010000K403 ..(KA's:
,
ANSWER: 088' (1.00) ? I k r
!
, ~,,~;... . - - .. , . _ , . . . . . .
-
- . , . , ,- _ , , , . . , , , . , , _ . _ , , , , , ,,.-,..,,,,._4., . . . , . . _
--
. . , , . ,
. . .__- _ _ _
_
,
I REACTOR OPERATOR () (). Page 8 i i I REFERENCE: l-l LO-LP-60301+07-C, pg. 12 ;
!
l KA 000007A103 [4.2/4.1) l l 000007A103 ..(KA's) l
!
! ! ! i i !
- ANSWER: 089 (1.00) .
j t l ) j REFERENCE: I VEGP 18007 C, pg. 2, 3, 5 & 11 l
h KA 000022G011 (3.3/3.6) : ! ! '
!
' 000022G011 ..(KA's) {
!
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- ANSWER: 090 (1.00) .i
* i i
! -REFERENCE: i l VEGP LO-LP-12101-30-C, pg. 35 .!
- VEGP 13011-1, pg. 2 ,
! KA 000025K101 [3.9/4.3] '
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000025K101 ..(KA's) F i F " a ANSWER: 091 (1.00) :
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* wwww y vyww we=~,, mw -w ....... REACTOR OPERATOR O O race ac REFERENCE:
LO-LP-60303-13-C, pg. 15-KA 000001K105 [3.5/3.8) 000001K105 ..(KA's) ANSWER: 092 (1.00) REFERENCE:
'
TS 3.4. KA 000009K320 (3.5/4.3) i
>
000009K320 ..(KA's) !
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.i ANSWER: 093 (1.00)
'l REFERENCE:
LO-LP 60303-13-C, pg. 30 KA 000003K309 {3.0/3.5} '
-000003K309 ..(KA's) ,
A. 'SWER : 094 (1.00) ,
.
^ '
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,#=--,.%- ,5,,- -',--s,,,--.,--t
.:.p-,,#.7,-y- . , . , , .w - . , , , , , - r -
- , ,, _,v.-,,ww.-<,.-.,cm,,--w-,--,. --
- . _ . . . _ _ . . - . _ _ _ _ _ _ . . _ . _ _ . - .-- - - - - - - - - - - - - - - . - .
.-.._.___.;
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REACTOR OPERATOR O O P 87 . t i REFERENCE: ;
VEGP LO-LP-60314-06-C, pg. 6
KA 000054K304'- - [4.4/4. 6) I
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000054K304 ..(KA's) i
! ANSWER: 095 (1.00) l !
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REFERENCE: l
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LO-LP 60321 07-C, pg. 9 ,
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l KA 000065A206 [3.6/4.2] ; 000065A206 ..(KA's) 1 .
.i ANSWER: 096 (1.00) ,
e '-! ] ! REFERENCE: ' I LO-LP 60305-04, pg. 6 ;
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KA 000015K102 [3.7/4.1) , 000015K102 ..(KA's) -! , ! , , , ANSWER: 097 (1.00) l ! I a r l'
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___ _ _ . _ _ . _ _ . _ . ~ . . . . . _ _ _ . . - _ _ . _ _ . _ . _ _ . - . _ _ _ . . . _ _ _ _ _ _ _ _ _ _ , _ _ _ _ _ . , REACTOR OPERATOR O O vase 88 ,
REFERENCE: f f
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VEPG LO-LP-28201-15-C, pg. 8 KA-000056A247 [3.8/3.9) F
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. 000056A247 ..(KA's)
1 ! t
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, ANSWER: 098 (1.00)
* t REFERENCE:
LO-LP-23301-11-C, pg. 10 KA 0000.r '02 (2.9/3.6) , e
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000036K302 ..(KA's) ANSWER: 099 (1.00) l REFERENCE: LO-LP-60320-04, pg. 7 KA 000051G011 (2.7/2.9) 000051G011 ...(FA's) 7.NSWER: 10 (1.00)
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' REACTOR' OPERATOR O O rese 89
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REFERENCE: f LO-LP-60323+02, pg. 5 i KA 000067G010 [3.3/3.7) f
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000067G010 . (KA's)
E
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REACTOR OPERATOR ()- Page 1
. ANSWER XEY
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i MULTIPLE CHOICE 023 d , 001 a 024 b 002 a 025 d ; 003 d 026 c -
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004 a 027 d
?
005 b 028 .. a 006 b 029 d ;
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007 d 030 d 008 d 031 a 009 d 032 , 010 d 033 c 011 b 034 d 012 c 035 d
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013 d 036 c 014 d 037 a , 015 a 038 d 016 d 039 a
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O l'/ b - 040 a 018 -b 041 d
. 019 c 042 a 020 a 043 b
--
021 d 044 a
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022 a 045 d
l ,
..
g p--ew +v -qy-yg-- 3 weay-ur'w-y y e * ,+eg Up w S+ e-.svgNt*-WrTTww+r-iw.e p yw g t M-gw33-9-- '- 'ww'wWW='y' -w3
. ~ - - - - - . - . - . . . . _ . - - - - - -
- - - - - - - .-.
..--.. - . - . - - .~ -.-_
i t
- REACTOR OPERATOR O O vase 2
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ANSWER KEY d
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f i
046 c 069 a j- ' 047 a 070 c l I 048 b 071 c
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049 b 072 a ; 050 c 073 c ' t
. l 051 b 074 b j
, 052 d 075 c 053 a 076 d I
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054 d 077 b ! l 055 c 078 d ;
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056 d 079 b i
.I 057 b 080 c i 058 c 081 d -
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059- a 082 d l 060 b 083 .a -
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-l 061 c 084 c
-
- 062 b 085 b 063 b - 086 b '
t 064 b 087 b
.; .. 065 c 088 c
! ! 066 a= 089 d
- 067 . b 090 c l
' 068 a 091 b
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-4
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'I WMP4"- T'W'WW M-WM N9MWW9'*- Wu rd w h- "#6'W"GN'W*b'** WWW N 'e T88"'erW w'e e Wee e.W'wat gireseter 87'tW~'*9**WUeeTWs sw 4 +5'"pt-S,Ter41h-hewe-war' hiw4v-+
. - . . - . .
REACTOR OPERATOR O O rase 3 ANSWER KEY ,
092 a i 093 d
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094 d 095 d 096 a 097 b 098 c 099 c
.100 a
, i i
(********** END OF EXAMINATION **********)
. () TESTCROSSREFERENCE() Page 1 (
t RO Exam PWR Reactor
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Organized by 0 u e s t:1 o n Number i OUESTION VALUE REFERENCE - 001 1.00 9000379 002 1.00 9000383 , 003 1.00 9000384- ! 004 1.00 9000386 005 1.00 9000387 006 1.00 9000388 007 1.00 9000390 008 1.00 9000392 : 009 1.00 9000393 010 1.00 9000394 011 1.00 9000395 ! 012 1.00 900039 .00 9000398 014 1.00 9000399 015 1.00 9000400 ; 016 1.00 9000401 ' 017 1.00 9000402-018 1.00 9000403 : 019 1.00 9000404 020 1.00 9000405 021 1.00 9000406 022 1.00 9000407 023 1.00 9000409 024 1.00 9000410 025 1.00 9000411 026 1.00 9000412 027 1.00 9000413-028 1.00 9000414 029 1.00 9000415 , 030 1.00 9000416' 031 1.00 9000417 032 1.00 '9000418-033- 1.00 -9000419 034 1.00 9000420 035- 1.00 9000421 036 1.00 9000422
:037 1.00 9000423-038 1.00 9000424 .;
039 -1.00 9000425 - 040 1.00 9000426-041 1.00 9000427 , 042 1.00 9000428 043 1.00 9000429-044 1.00 9000430 045 1.00 9000432 046 .1.00 .9000433 047 1.00 9000434-04 .00 9000436-049 1.00 9000437 ,
-r n- -- . i._._.._.__.- .-.,_,;__-,;...,___.-_, ..;_;--.--...-, , - - , . . - ,_ 9
... . . . - . . - . - - . - _ _ . - . . _ - _ . _ . . _ . . - . _ . . . - . - . . - _ . . - . . . . () TEST CROSS REFERENCE ( )' Page 2 RO Exam PWR Reactor 1 0rgani2 ed by Question Number l
J ' i l OUESTION VALUE REFERENCE ; 050 1.00 9000438 1 051 1.00 9000439 052 1.00 9000440 053 1.00 9000441 ' 054 1.00 9000442 , 055 1.00 9000443 056 1.00 9000444 057 1.00 9000445 058 1.00 -9000446 059 1.00 9000447 060 1.00- 9000448 i 061 1.00 9000449 062 1.00 9000502 063 1.0 .00 9000452 065 1.00 9000454 066 1.00 9000455 067 1.00 9000457 068 1.00 9000459 069 1.00 9000460
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070 1.00 9000461 071 1.00 9000462 072 1.00 9000464 073 1.00 9000465 074 1.00 9000467 075 1.00 9000469 . 076 1.00 9000470 f 077 .1.00 9000471 07 .00 9000472 ' 079 1.00 9000474-080 1.00 9000476 081 1.00 9000478 082 1.00 9000479
083 1.00 9000480 084 1.00 9000482 085 1.00 9000483 ; 086 1.00- 9000484_ t 087 1.00 9000485 088 1.00 9000486' i 089 1.00- 9000487 090 -1.00 9000488 091- 1.00 9000409 092 1.00 9000490-093 1.00 9000491-094 1.00 -9000493' 095 1.00 9000494 096 1.00 19000495-097 1.00- -9000499 098 1.00 9000500
- .a . . . . _ - :- - - . . , . _ _ - -.- ...:..-. . . . . . . ,, - - , - , =
. _ . .__ ._
_ _ . - - _ _ _ _ _ _ _ _ _ _ -.
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h TESTCROSSREFERENCh Page 3 ' l RO Exam PWR Reactor '
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Organized by 0uest i o'n Number- ! s
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l -l i ! OUESTION VALUE REFERENCS l j- i l 099 1.00 9000501 !
' .00 9000503 j' ......
!
- 100.00 1
......
- ...... )
] 100.00 l
f 2.
. -I h l b 4 s
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--- . w.vemM+m E.w _ ._ _ ww w -- _w.rm-.--a. . - = . - - . - - - - - . . _
_ _ _ Page 4 TESTCROSSREFERENC( RO Exam PWR Reactor Organized by KA Group PLANT WIDE GENERICS QUESTION VALUE KA 005 1.00 194001A103 007 1.00 194001A103 006 1.00 194001A111 008 1.00 194001K101 004 1.00 194001K101 003 1.00 194001K102 - 002 1.00 194001K102 013 1.00 194001K103 012 1.00 194001K103 009 1.00 194001K103 011 1.00 194001K104-010 1.00 194001K104 001 1.00 194001K105
......
PWG Total 13.00 PLANT SYSTEMS Group I QUESTION VALUE KA 014 1.00 001000A102 016 1.00 001000A106 - 015- 1.00 001000K402 017 1.00 001000K504 019 1.00 003000G015 018 1.00 003000K201 020 1,00 003000K201 024 1.00 004000A207 023 R1.00 004000K104 021 1.00 004010A204 026 1.00 013000A301 047 1.00 013000A402-025 1.00 013000K103 046 1.00 015000A3C3-044 -1.00 015000K301 027 - 1.00- -017020A201 028 1.00 022000A301-029 1.00 059000X419 032 1.00 OG1000A102 031 -1.00 061000K103-030 1.00 061000K404 034 1.00 072000G008'
......
-PS-I Total 22.00 d
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() TESTCROSSREFERENCE() Page 5 RO Exam PWR Reactor Organized by KA Group PLANT SYSTEMS Group II QUESTION VALUE KA 039 1.00 002000K106 040 1.00 002000K402 038 1.00 006000G004 022 1.00 006000K603 -
036 1.00 010000A403 060 1.00 010000G013 087 1.00 010000K403 035 1.00 010000K403 037 1.00 011000A211 049 1.00 012000G008 , 048 1.00 012000K401 045 1.00 014000G015 042 1.00 033000K303 033 1.00 062000A401 041 1.00 063000K301 053 1.00 064000A101 054 1.00 064000A406 052 1.00 064000G007
......
PS-II Total 18.00 Group III QUESTION -VALUE KA - -- 056 1.00 005000K201 059 1.00 007000K401 058 1,00 008000A301 061 1.00 008000K301 057 1.00 008030A304 050 1.00 027000A401 051 1.00 020000G005 043 1.00 034000A201 055 1.00 103000A101
......
PS-III Total 9.00
...... ......
PS. Total 49.00
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(]) - TESTCROSSREFERENbE(] Page- GL r RO Exam P WLR Reactor
- O'r g a n i-z e d K A-
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F- buy G r oz up , RGENCY PLANT-EVOLUTIONS Group I-QUESTION VALUE KA 081 1.00 000015G010 096 1.00 000015K102 082
,
1.00 000024K301-084 1.00 000026K303 076 1.00 000040K101 075 1.00 000040K104 085 1.00 000051A202 099 1.00~ 000051G011 071 1.00 000055G012 069 1.00 000055K302 062 1.00 000057A217 086 1.00 1 000067A105 100 1.00 000067G010 065 1.00 - 000069G012 0C8 1.00 000074K208-EPE-I Total k5bb Group II QUESTION VALUE KA-Ous 1.00- 000001G003 091 1,00 000001K105 093 . 30 000003K309 088 . J' 000007A103 070 5.00 000007A202' 073 1.00 000008K302 066 1.00 000009G012 092 1.00 000009K320
,077- 1.00 000009K323 064 11.00 000011G010 078 1.00- 000011K305-079 1.00 000011K313 089 1.00 000022G011 090 1.00 - 000025K101 072' 1.00 - 000029A209'-
083- 11.00 000038A144
080 1.00 000038K306-094 1.00 000054K304 074 1,00' 000054K305
......
EPE-II Total -19.00
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() TESTCROSSREFERENCd) Page 7 RO Exam PWR React ar Organi ed by KA Group EMERGENCY PLANT EVOLUTIONS Group III QUESTION VALUE KA 098 1.00 000036K302 067 1.00 000056A203 097 1.00 000056A247 -
095 1.00 000065A206
......
EPE-III Total 4.00
...... ......
EPE Total 38.00
...... ...... ......
Test Total 100.00 s
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NRC Official Use Only ,
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j _ VoyHe L if:/ si..r k E ra n. 9 z- 3 o z. - . 8g Nuclear Regulatory Commission Operator Licensing Examination This document is removed from Official'Use'Only category on-date of examination.
l , ! NRC Off.icial Use Only-L
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_.il Oi O U.=S. NUCLEAR REGULATORY COMMISSION SITE SPECIFIC EXAMINATION-
-SENIOR REACTOR OPERATOR LICENSE REGION lr CANDIDATE'S NAME:
FACILITY: Vogtle-1-& 2-REACTOR TYPE: PWR-WEC4 DATE ADMINISTERED: 92/12/10 _ 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 questio The passing grade requires a final grade of-at least 80%. Examination papers will' be picked up -four (4) hours-after the-examination start CANDIDATE'S-TEST VALUE SCORE % 100.00 % FINAL GRADE All work done on this examination is my ow I have neither given.nor received ai Candidate's Signature-
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SENIOR' REACTOR OPERA : -
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ANSWER S H ELE-T Name: Multiple 1 Choice (Circlelor 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 -ir 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 03 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 d
'013 a b- c d 036 a b c d-014 a b c d- 037 a' -b - c d 015- a b c d 038 -a b c d 016 a b c d 039- a- b c d 017 a b c d- 040 a =b c- d 018 a b c d 041 a b c- d 019 .a b- c d 042 a b c d 020 a1 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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. SEN OR' RFACTOR _ OPERAT -P ge -3-
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ANSWER SHEET- -i l Name:
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Multiple' Choice (Circle or X your choice)
-.If you change your answer, write your selection in the blan _+
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046 a b c d 069 a b1 c: .d i 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 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 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 d58 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 MATCHING-064 a b c' d a-065 a -b c .d a b c d MULTIPLE CHOICE i.-
~ -067 a b c a- .b c d 068 a b= - d 088 a- c .d , , . - . - . , . - .- .- -- , .:.-.-
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SENIOR REACTOR OPERA . .: . Pag e 4'- A N S W E: R .S H.E'E !
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Name: Multiple Choice (Circle or X your chcice)c If you changelyour answer, write your selection in the blank.
089 a b c d 090 a b c d 091 a- b c d 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
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(********** END OF EXAMINATION **********)
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page-- 5 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
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and could result 1in more severe penaltie . 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 examinatio . Restroom trips are to be limited and only one applicant at-a time may leav You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheatin . Use black ink or dark pencil ONLY to facilitate legible reproduction . Print your name in the blank provided in the upper right-hand corner of the examination cover sheet and each answer shee . Mark your answers on the answer sheet provide USE ONLY THE PAPER PROVIDED AND DO NOT NRITE ON THE BACK SIDE OF THE PAG . Before you turn in your examination, consecutively number each answer sheet, including any additional pages inserted when writing your answers-on the: examination question pag . Use abbreviations only if they are commonly used in facility _ literatur Avoid using symbols such as < or > signs to avoid a simple transposition error resulting in an incorrect answer. Write it ou . The point value for each question is indicated in parentheses after the questio . Show all calculations, methods, or assumptions used to obtain an answer 1to _ any short answer question . Partial credit may be given except on multiple choice questions. Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLAN . Proportional grading will be appl e Any additional ~ wrong information that-is provided may-count against yo 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 .
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0.20 point If one of your five responses is incorrect, 0.20 willabe deducted and your total credit _for that1guestion will be 0.80 instead-of-
-1.00 even though you1got the-four correct answer . If the intent of a question is unclear, 'ask questions of' the examiner only.
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Page; 6-14. When-turning;in your examination, assemble:the_ completed examination witf[ examination questions, examination aids and answer sheets'. In addition, turn in all-scrap. pape . Ensure all information you wish~ to have evaluated as part of your answer is
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on your answer sheet. Scrap paper will be disposed of immediately; following the examinatio . To pass the examination, you must achieve-a grade of 80% or greate . There is a time limit of four (4) hours for completion of the examinatio . When you are done and have turned in your examination, leave the examinatio 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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SENIOR ~ REACTOR OPERAT - Page-_:7
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QUESTION: 001 (1.00) When restoring _a FAIL CLOSE air-operated valve -(AOV) that has a handwheel, after it has been used as a closed boundary valve for a clearance, which one of the following correctly descr1bes the1 order (first to' last). for removing the hold tags? a. Handwheel,. valve air supply, handswitch b. Handswitch, handwheel, valve air supply c. Valve air supply, handswitch, hand'vheel d. Handwheel, handswitch, valve air supply OUESTION: 002 (1.00) Following repair of a damaged-breaker, which clearance status is required for breaker testing to be performed? a. All hold tags must-be temporarily cleared, the breaker racked to'- the test position, and a hold tag-placed on the racking devic b. Necessary hold-tags must remain on1the breaker, the_ breaker racked to the-test position, and a caution tag placed on the- , racking device, c. All hold tags must be removed and' replaced with caution _ tags, and the breaker racked to the-test positio d. Necessary hold tags must be removed via a. functional ~ release, and the breaker racked to the test position.
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J SENIOR REACTOR OPERA - _PageL'8-
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QUESTION: 003 (1.00)
.Which one of the following describes.how an independent verification of-- -the position of a manually operated THROTTLE valve-is-performed?
a. Compare a visual observation of the-stem position or indicator-position with the-required position, b. Completely close the valve and then open it to the required position, c. Completely open the valve and then close it to the required positio _ d. Move the valve exactly one turn _in the closed direction and then return it one turn to its required positio QUESTION: 004 (1.00) The USS must authorize surveillance tests.by signature if the test: a. affects both units, b. manipulates plant equipmen c. is required for a mode change, d. has less than a 72-hour Action Statemen QUESTION: 005 (1.00) Per VEGP 10000-C, " Conduct of Operations," which-one of the following_ may NOT be adjusted by shift operating personnel during at-power operations? a. Power-range gain adjustment following a-calorimetri b. Intermediate range compensating voltage settin c. Data A/B-selector switch behind the DRPI display, d. Atmospheric relief. valve setpoin _ - h disimi-- i i .I- _ _ _ . . _ _ _ _ _ _ _ _ _ _ , _ _ _ _ _ _ _ , _ _
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- QUESTION: ~006 '(1.00)
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: Emergency exposure limits are rem.'whole' body for_ life saving activities and -(2) rem'whole body exposure for corrective or '
. protective: activitie . (1) 50; ( 2 ) - 10 - ^-
d (1) 50; (2) 25 (1) 75; (2) 10 (1) 75; (2) 25
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QUESTION: 007 (1.00) In accordance with ALARA guidelines, Independent Verification of components is not required--if significant= radiation exposure is involved. Which one of the following is the LOWEST whole-body dose that i would constitute significant radiation-exposure?
- - a. 5 mrem- mrem -
c. 20 mrem d. 25 mrem
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- SENIOR REACTOR OPERATO . Page 10 QUESTION: 008 .(1.00)
Which one of the:following current whole-body exposures ~will cause a "NO ENTRY" display on the RCA. access point'HP computer terminal upon an
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attempted log-in? Where applicable, assume an updated Form 4 IS on file,-lower dose-limit for pregnancy have NOT been chosen, and individual-has.NO exposure-history other than that liste a. Male temporary radiation worker with 50 mrem in current quarte b. Female radiation worker with 350 mrem in current: quarter, c. Male radiation worker with 850 mrem in current quarter, d. Female. radiation worker with 4150 mrem in current yea QUESTION:-009 (1.00) An individual's'TLD shall be processed.before the individual.is authorized.to receive greater than: a. 1000 mrem in a quarte b. 1500 mrem in a quarte c. 4000 mrem in a year, d. 4500 mrem in a yea _ , s
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; QUESTION: 010: -(1.00)
Given the following. conditions at Unit 1:-
- Reactor Power at 10%, startup in progress -
Turbine--load being increased slowly-If rod control is inadvertently placed in AUTO under these conditions, which one of the following describes the response of the control rods?
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a. The control rods will NOT move out until Tavg is at least' degrees F greater than Tre b. The control rods will move out at-the rate of 8 steps per minute to keep Tavg and Tref within 1.5 degrees c. The control rods'will NOT move until CB D is manually positioned-to 12 step d. The control- rods will NOT move out until Turbine Impulse Pressure (PT-505) exceeds 15% powe QUESTION: 011 (1.00)
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Which one of the following describes the parameter (s) which the RIL computer uses to generate a low rod position setpoint? a. Tavg and Tref, b. Core delta-T, ' c. Tref and core' delta- d. Tref and-controlling group rod height.
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. QUESTION: 012 (1.00)
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Which one of the following statements describes-the basis for each ' reactor coolant pump - (RCP) . having a Class 1E breaker in series with its-non-1E breaker?-
a. To provide' electrical isolation from the-associated 13.8 kv bus during motor maintenance and circuit breaker testin b.'To prevent a mechanical failure of the-containment' electrical penetration due to' fault current if the non-1E breaker fails, c. To prevent damage to the supply breaker _forlthe associated-13.8 kv bus in the event of an RCP locked roto To prevent damage to the supply breaker for the; associated 13.8 kv bus in the event of a seal failure grounding the moto QUESTION: 013 (1.00) Which one of the following is the electrical bus which supplies power to-Reactor Coolant Pump (RCP) 3 at Unit 1? NAA.
. b. 1 NA AA02.
d. 1BA03.
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SENIOR REACTOR OPERAT _ P' age 13 LQUESTION:-014 -(1. 0 0 )
-. Unit 1Lis-operating at 100% power at BOL w.th all syscems operable when a complete -loss of instrument air occurs. Which one of the following i lists _ Chemical and Volume Control System (CVCS) valves that_will-fail OPEN?
a. Charging Flow Control Valve (FV-121), RHR Cross Connect Valve (HV-128), b. Letdown Leak Protection Valve (HV-15214), Auxiliary Spray Valve (HV-8145).
i c. Orifice Isolation Valves (HV-8149A, B, C), Containment Isolation " Valves (HV-8160, 8152).
d. Seal-Injection Flow Control Valve (HV-182), Letdown Pressure: Control Valve (PV-131).
QUESTION: 015 (1.00) During operation at 100% power at Unit 1, power is lost to the_ motor operator on valve.HV-8112,. seal return line-containment isolation. valv Which'one of_the following describes the effect on RCP seal-return parameters of the loss of power to HV-8112? , a. Pressure will slowly rise until relief. valve PSV-8121 opens, b. Pressure will increase but the check valve around HV-8112 will open preventing relief: valve _ actuatio c. Temperature will increase slowly due to reduced flow-and may-require-RCPs to be stopped, d.-Temperature and pressure will remain normal because valve 1HV-8112 will not. change _ positio ;
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3 QUESTION: 016 '(1.00) Which one of the following-is-the minimum required normal charging header flowrate if an Emergency Boration is to be performed from the ; Boric Acid Storage. Tank (BAST)? l H a'. 30 gp b. 42 gpm c. 87 gpm gpm QUESTION: 017 (1. 00 ) Which one of the following will NOT' result in the generation of a Containment Ventilation Isolation (CVI) signal on a high alarm? a. Containment area radiation low range monitor RE-000 b, Containment vent particulate monitor RE-2565A . c. Containment vent iodine' monitor RE-2565 d. Containment atmosphere particulate monitor RE-2562 QUESTION: 018 (1.00)- With the Solid State Protection System,(SSPS). Train A Mode Select'or Switch in-" TEST",_what is the condition of Train A SSPS?
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' Input relays are-prevented from providing a " tripped"-inputLto-the logic card, b. The logic card is prevented from recognizing.any blocks or- _s permissives that are activ ,
c.-The Spray Test and Output' Relay Test _ circuits are enabled, d. Train A SSPS is still able to actuate ESF equipment on a valid ESF initiation signa , k -+ , -- - _ . _ _ _ - - . . . - - -, y w ., , , .-
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SENIOR REACTOR'.OPERAT Page 15-
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QUESTION: 019 ('1. 0 0 ) t t Which.one of the following describes the indication which would.be lobtained if a. thermocouple circuit suffers a short in the junction box?s a. Temperature indication would be off scale low, i b. Temperature indication would be off~ scale hig c. Temperature indication would oscillate at.the high en d. Temperature indication would be that of the junction bo , r QUESTION: 020 (1.00) Which one of the following describes the response of the containment cooling fans to a safety Injection signal?. a. Fans running'in fast speed will stop; then all fans will-start' in slow speed at'30.5 second .
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b. Fans-running in fast speed willistop;-then fans 3, 4, 5-and 6-will start in slow at 30.5 seconds, and fans 1, 2,:7-and.8 will'- _ start: in slow af ter 50.5 seconds, c.. Fans running in fast speed will shift'to slow-speed; then the-idle fans will sequence on in slow speed after 50.5 seconds, d. Fans running in fast speed will continue operating.until--'5 seconds; then all fans will start / shift to slow speed ~. N
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F QUESTION: -021 (1.00)-
' 'Given the following conditions at-Unit 1:
_-A reactor trip has just occurred from 25% power-
- No safety injection signals are present ; -
Tavg is 563 degrees F, decreasing slowly ,
- All steam generator levels are greater than 27% narrow range Which one of the following describes the response of the main feedwater
_ system? a. Main feed regulating valves and bypass feed regulating valves close; MFPs trip and the main steam lines isolat b. Main feed regulating valves, bypass feed regulating valves, and bypass feed isolation valves close; and MFPs tri c. Main feed regulating valves and bypass feed regulating valves close; and the main steam lines isolate.
d. Main feed regulating' valves, bypass feed regulating valves, and-bypass feed isolation valves close.
, QUES :ON: 022 (1.00) Which one of the following describes the mechanism by which adequate flow is assured to intact steam generators (SGs) in the event of an AFW feedline break?
- a. Capacity of the AFW pumps, b. Sizing of the AFW feed line " Quick acting" isolation valves in AFW lines to SG d. Flow restrictors in AFW lines to SG ~ ~
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; ~ ; SENIOR: REACTOR'OPERAT Pagei17 2-QUESTION:-_023- (1.00)
Which one of the following describes the source of steam to-the turbine-- driven auxiliary feedwater pump? , a. SGs 1 and 2
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b. SGs 3 and 4 c. SGs 1 and 3 d. SGs 2 and 4
' QUESTION: 024 (1.00)
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What is the loweet steam pressure at which the turbine-driven AFW pump can be operated? a. 30 psig, psi c. 90 psi ' psig.
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' QUESTION: 025 L(1.00)
Given the following conditions at-Unit 1:
- Reactor power-at 354
- Non-1E buses have been transferred to the UATs
- All buses _are energized -
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- No indicating lights displayed on high side circuit switcher for
'the RAT-Which one of the following describes the condition that would cause these indications to exist?
a. The RAT high side circuit switcher _has_ auto trippe b. A fault exists in the RA ' c. Control power is lost, the RAT nigh 015e circuit switcher is close d. The fast bus transfer to the RAT has actuate QUESTION: 026 -(1. 0 0 ) Which-one of the following describes the effect of decreasing the- . , Pressurizer Pressure Master Controller _ potentiometer setting to 2185 psig during Mode 1 operation? a. Actual system pressure will remain between 2250 and_2260:psig but-the low pressure alarm will actuat b. Actual system pressure 1will remain in the normal; control band of-2250-2260 psig, but.the PORVs will' receive a block signa c. Actual system pressure will decrease.to 2185 psig, and the PORVs
- will receive a block signal, d. Actual system pressure will-decrease to 2210'psig, and then the-backup heaters will raise pressure to 2218 psig.
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ISENIOR. REACTOR-OPERAT6 - Page'19 .
-QUESTION: _027 :(1.00); :
I Which one of the following describes an accident which the-ECCS is designed'to mitigate? -
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a. Loss offelectrical powe b. High reactor power-transien c. High startup rate in the power rang ., d. Feedwater line brea QUESTION: 028 (1.00) Which one of the following describes one or more penetrations in loop 3 cold leg, in-addition to the accumulator / safety injection /RHR and BIT connections? CVCS letdown, b. Normal Charging, PZR spray, c. CVCS excess letdown, sample line, d. Alternate Charging, PZR spra QUESTION: 029 (1.00) Which one of~the following describes the-. locations where reactor-vessel '
. level-(RVLIS) indications:are'available? -
a, Plant Safety Monitoring System computer and the Emergency Response Facility (ERF) computer.
' b. QMCB recorder and local display gage c. Proteus computer and the Technical. Support-Center compute d. QPCP level indications and ERF compute _ . - . . . -
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P$ge-20 QUESTION: 030 (1. 0 0 ) .
; The Control Room annunciator for Spent Fuel Pool High Temperature is out of service. _ Which one of the following would be an alternate indication-of a loss of Spent Fuel Pool Cooling?
a. Local audible high--temperature alarm in the Fuel Handling Buildin > Local audible alarm of Spent Fuel Pool Cooling High Temperature on the-Liquid Waste Processing Panel, c. Spent' Fuel Pool High Humidity clarm in the Control Room, d. High alarm on RE-008, Fuel Handling Building Area Radiation Monito QUESTION: 031 (1.00)
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Given the following conditions at Unit 1:
- Reactor startup is in progress - All NI switches are=in their normal lineup - No manual blocks have been inserted - Intermediate Channel N35 indicates 2E-10 -
Intermediate Channel N36 indicates 9E-11'
- Power is lost to Source Range Channel N31 - Power is maintained to Source Range Channel N32 Which-one of the following describes the response of.the reactor. plant?
a. A reactor trip signal is generated-resulting in a reactor tri b- . A reactor trip signal =is generated but no trip occurs 1since one channel is above P- c. No reactor trip signal is generated since one channel is above _ , P-6.
I d. No reactor. trip signal is generated -but the level- trip switcli must be taken to bypass as soon.as'N36 indicates greater than - 1E-10.
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LQUESTION: 032 . (1.' 0 0 )
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Given the~following conditions at Unit 1: -1
- Reactor startup is in progress - - Source: range channel N31-indicates 1E5 - ' Source range channel _N32 indicates'9.5E4- - Intermediate range channel N35 indicates 4E-10 - Intermediate range channel N36 indicates 2E-11 .
Which one of the-following statements describes the condition of the-
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nuclear instruments? a. N35 is overcompensate b. N35 is undercompensate ' c. N36 is overcompensate d. N36 is undercompensate . QUESTION: 033 (1.00) Given the following conditions-at Unit 1:
- Plant is in Mode 3 - ~
RCS. pressure is 750 psig
- RCS temperature is 400fdegrees F - Main steam line isolation signal has been actuated and not: reset - Steamline SI and Pressurizer SI have been blocked:
While attempting to reset the-main steam-line-isolation signal,-the
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operator mistakenly resets.the low steamline pressurelSI. -Which1one of- -
-the following-describes the automatic responseLof the-plant?-
a. .Jul SI signal will' be generated and an SI will occur, b.-An SI signal will be generated.but, an SI'will NOT occur since
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the' plant is in MODE c.-An SI signal will NOT be generated since the' plant is in MODE;3, d . 1u1 SI signal will NOT'be generatedesince the: plant is-in. MODE 3-
- --- and-the MSLI-is actuated. -
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TQUESTION: 0343 ( l '. 0 0 )
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Given'the_following. conditions at Unit 1:
- - Reactor _-in Mode 3 - Preparations being made for reactor-startup- - SSPS Train A Mode Selector Switch is in TEST - Input Error _ Inhibit Switch is in INHIBIT --Trip Breaker A is closed - Trip Breaker B is racked out for testing-If the operator attempts to close Bypass Breaker B for testing, which one of the following describes the response of the plant?- -
a. The General Warning alarm for Train B will actuate and inhibit any reactor trip signal b. A reactor trip signal will be generated when Bypass Breaker B is-closed,
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c. The multiplexer will shift to Train B until Logic-Train A is returned to " Normal".
d. Train B slave relays will align for bypass breaker operation QUESTION: 035 (1.00) Which one of the following will result in illumination of the "PCS: CABlNETS PWR SUPPLY FAILURE" annunciator for'the Reactor Protection
- -System (7300) Process Cabinets?
a. Loss of the NYRS electrical bu b. Loss of 24 VDC backup power.to NSSS protection cabinet ~ c. Energizing the Master Test Relay'before placing the bistable
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test switches in TES d. Loss of power to the' Eagle 21 cabinet.
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; QUEC"LN: 036. (1.00)
Which one of-the following describes CONTROL ROOM indications that the Containment Iodine Removal System is in operation.following a LOCA inside containment? a. Containment spray pump suction pressure and discharge pressure
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at nominal operating values with both spray additive tank isolation valves ope Containment spray actuation signal present, at least!one spray-additive tank isolation valve open, and containment spray pump discharge flow rate at nominal operating value, c. Spray additive tank level decreasing, containment-sump 'avel increasing,-and containment spray actuation signal present, Containment pressure decreasing, containment h".nidity increasing, and containment sump level increastn QUESTION: 037 (1.00) Which one of the following is the upper limit on hydrogen concentrittion in containment for starting a hydrogen recombiner? a. 4 %. %. %.
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SENIOR REACTOR OPERAT Page 24 QUESTION: 038 (1.00) Given the following conditions at Unit 1:
- A LOCA has occurred
- Following the LOCA there was a fault on the P.eserve Auxiliary Transformer (RAT)
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"A" Emergency Diesel Generator (EDG) has started and is supplying its Emergency Bus
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"A" EDG control air system piping has just ruptured Which one of the following describes the response of the "A" EDG to the loss of diesel control air pressure?
- The engine will continue to run, but the output breaker will open causing the engine to trip on overspee b. The engine will stop and the output breaker will open deenergizing the Emergency Bu c. The engine will stop, the output breaker will remain closed, and the generator will motoriz d. The engine will continue to run, but all engine protective features will be inoperabl m a
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. . SENIOR REACTOR OPERAT Page 25 QUESTION: 039 _(J .00)
Given the following conditions _at Unit 1:
- Reactor Trip and Loss of Site Power has occurred-
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"A": Emergency Diesel-Generator (EDG) starts, but output breaker fails to close
- Bus 1AA02 is deenergized
- Crankcase pressure indicates 3 psig
- Lube oil pressure.is 25'psig
- Jacket water temperature is 200 degrees F
- Two minutes after starting, the "A" EDG shuts down
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Which one of the following caused the "A" EDG to shut down? - Lube oil pressure, b. Crankcase pressur c. Failure of output breaker to close, d. Jacket water temperatur QUESTION: 040 (l'. 0 0 ) An: Emergency Diesel Generator is started for a test and.is run for 48 hours at loads between 1500 KW and 3000 -}G1. Which one-of the following_ describes the action required to be taken at the conclusion of the 48 hour test run? _ a. Review all-readings and trends;;if satisfactory, shut down th _ engine, b. If any cylinder exhaust temperatures havd_ exceeded 900_ degrees = __F,_run at 3000 KW for an additional 2 hours,-and then stop1the engin Idle the engine unloaded _for-2 hours or_untilitemperatures (lube
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oil and - cooling water)' stabilize,- and then stop the engin d.-Run the engine.at 3500 KW-load for 4 hours, and then shut down the engin l _ _ _ _
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___ SENIOR REACTOR OPERATL Page 26 OUESTION: 041 (1.00) Which one of the f ollowing lists the power supplies f or the RilR pt.mps? a. INA05 and INA04.
l b. 1NAA and 1 NAB.
c. 1AB05 and 1BB07 d. 1AA02 and 1BA0 QUESTION: 042 (1.00) Which one of the following describes the response of the CCW system if pumps, 1, 2, 5 and 6 ara running when a Safety Injection is MANUALLY initiated? a. All pumps will stop; pumps 1, 2, 3 and 4 will start at the appropriate sequencer ste b. Pumps 1, 2, 5 and 6 will continue to run; pumps 3 and 4 will start n' -he appropriate sequencer ste c. Pumps 1, 2, 5 and 6 will continue to run, and no others will star d. Pumps 1 and 2 will continue to run, but pumps 5 and 6 will stop when pumps 3 and 4 sequence on.
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SENIOR REACTOR OPERA' Page 27 QUESTION: 043 (1.00) Given the following conditions at Unit in > - Reactor in Mode 4
- Cooldown to Mode 5 in progress Train A CCW Pumps 1 and 3 are running 1
- Train A CCW Pump 5 in standby I
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If a spurious CCN Surge Tank Lo-Lo level is detected by the circuit that I controls CCW Pump 1, which one of the following describes the response of the Train A CCW Pump (s)? a. CCW Pumps 1 and 3 will continue to run until a second lo lo . surge tank level is actuated (a 2/3 coincidence).
l b. CCW Pump I trips off; CCW Pump 3 continues to run, but-CCW Pump i 5 will not autostart due to the lo-lo surge tank level signa ; c. CCW Pump 1 trips off; CCW Pump 3 continues to run, and CCW Pump' t 5 autostarts on low. discharge header pressur i d. CCW Pumps 1 and 3 trip off and CCW Pump 5 is provented from .
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autostarting due to the lo-lo surge tank level signa :
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QUESTION: 044 (1.00) Which one of the following best describes the methods used to cool the - Pressurizer Relief Tank (PRT)? * a. Recirculation through Reactor Coolant Drain Tank (RCDT) heat exchanger or spray f rom. Reactor Makeup . Water System ' (RMWST) . b. Recirculation through RCDT heat exchanger or spray from Residual Heat Removal (RHR) recirculatio c. Spray from RMWST or venting to Waste Gas Decay Tank - _, d. Spray from RHR recirculation or spray from excess letdow L
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l SENIOR REACTOR OPERAT - Page 28 QUESTION: 045 (1.00) During recovery from a small break LOCA, which one of the following actions will reduce the volume of a void formed in'the vessel head? a. Operate Pressurizer spray ) i b. Operate Pressurizer heater ) c. Stop all Reactor Coolant Pump ) d. Stop Safety Injection Pump i
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QUESTION: 046 (1.00) The following plant conditions exist:
- Unit 2 is in MODE Reactor water level is 22 feet above the vessel flang , - RHR pump B is operating; RHR-pump A is operable, but stoppe The f ollowing annunciators have just alarmed: + CCW TRAIN A LO HDR PRESS annunciator , - CCW TRAIN A LO FLOW annunciator - CCW TRAIN A RHR PMP SEAL LOW' FLOW annunciator j - CCW TRAIN A RHR HX LO FLOW annunciator The Aux Building Operator reports flowing water, indicating a pipe-rupture has occurred somewhere in the CCW pipin ,
Which one of the following is a required action for these conditions? Increase level in the Train B CCW. surge tank to 65 percen b. Reduce level in.the Train A CCW surge tank to 35 percen , c. Increase reactor water level to-23-feet'above the vessel-flang d. Reduce reactor water level to 21 feet above the vessel. flang , - . . = . _ _ ~ , . _ . . - _ , - . . ~ - . . . . . . . - - - _ ---,.-,4,~-..-~,-.,.-,.- ..--,_,,._-~.,,,--6,,-,- _,,.--,;
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I QUESTION: 047 (1.00) : . The heat flux hot channel factor limit calculation includes the term K(z), which is a normalized peaking factor that varies as a function of core heigh . Which one of the following is the reason K(z) must be used in the heat flux hot channel factor limit calculation? : a. Compensates for the increased coolant temperatures that occur at higher core height b. Adjusts for the longer time delay in core reflood at higher' core heights following a LOC c. Adds a conservative uncertainty factor since core flow becomes
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increasingly turbulent at higher core height d. Accounts for greater power production in the upper regions of the core near EOL due to axial flux shiftin QUESTION: 048 (1.00) i One minute after a coincident reactor overpower accident /large-break-LOCA occurred, containment pressure was noted to be steady at 12 psig and containment radiation was noted to be steady at.2E+5 rad /h ,
" .Therefore, the use of Adverse Containment Conditions parameters was directed by the USS.
> Which one of the following sets of conditions would allow discontinuing the use of Adverse Containment Conditions? CURRENT CURRENT CURRENT CONTAINMENT PRESSURE ' RADIATION LEVEL INTEGRATED RADIATION (PSIG) (RAD /HR) LEVEL (RAD) .0 -1.5E+5 3.5E+5 .5 1.5E+4 1.5E+5 .5 SE+4 1.5E+6 .0 SE+3- 6.5E+4 i l
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_ . . . . SENIOR REACTOR OPERA - - Page 30-QUESTION: 049 (1.00) An EOP step reads as follows: Verify FW isolation: o MFIVs - SHUT o BFIVs - SHUT o MFRVs - SHUT o BFRVs - SHUT The LULLETS ("o") indicate thats a. the valves must be checked in the specified sequence (MFIVs; - then BFIVs; then MFRVs; then DFRVs).
b. closing of these valves is more important than other actions without bullet c. the valves must all be checked, but the order in which they are checked is not important, these valves should already be in the correct position, assuming
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that automatic actuations have occurred correctl QUESTION: 050 (1.00) With RCS pressure currently greater than 1375 psig, the following RNO-is encountered while performing procedure 19010-C, " Loss of Reactor or Secondary Coolant,"
"IF RCS pressure lowers to less than 1375.psig, THEN stop RCPs."
How long is this conditional action directive applicable? a. Throughout 19010-C only, b. Only-during this step of 19010- c. Throughout the entire EOPs (FRGs and ORGs).
'd, Throughout 19010-C and after any transition to another'OR . . . .
-- SENIOR REACTOR OPERAT Page 31 . QUESTION: 051 (1.00) Which one of the following describes how REINITIATION of safety injection (SI) will occur if offsite power is lost following an SI reset? a. SI will occur automatically upon the loss of offsite powe b. SI will occur only if manual action is taken by the operator, c. SI will occur automatically when required SI' actuation conditions exis d. SI can occur manually or automatically following the load shed - by the sequence QUESTION: 052 (1,00) In the post-accident monitoring system, the RVLIS hydraulic isolators are designed to: a. maintain the containment pressure boundary in the event of a-level sensing line brea b. adjust pressure in.the level sensing lines to accommodate changes in containment pressur c. regulate flow to the transmitters to prevent damage from reactor pressure and level change d. retain operability of RVLIS in the event of a loss of_ power or instrument ai . . .. .
SENIOR REACTOR OPERA - Page 33 l l
l QUESTION: 053 (1.00) l l One of the major directives contained in VEGP 19100, " Loss of All AC l 4 Power," is to: a. perform secondary depressurization in a rapid manner even if , pressurizer level is lost during the depressurizatio l refrain from resetting any SI signal to allow the SI sequencer to actuate when bus power is restore c. enable the autostart of all ESF motor loads to ensure rapid recovery upon power restoratio d. complete each step of the procedure prior to moving on to the i next ste QUESTION: 054 (1.00) Following a Reactor trip on Unit 1, VEGP 19000-C, " Reactor Trip or Safety Injection," is entere When step 3 is reached, it is determined that 480 Vac busses 1AB04 and 1BB16 are NOT energize Which one of the - following describes the actions required at this point? i a. Enter VEGP 19100 C, " Loss of-All AC Power," since one emergency bus on each train is deenergize b. Stop VEGP 19000-C at step 3 until the'deenergized busses are reenergized from the diesel generato c. Continue with VEGP 19000-C while attempting to restore power to
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the deenergized busses, d. Continue with VEGP 19000-C; no action is required for the deenergized busse !
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SENIOR REACTOR OPERA . _Page 33-QUESTION: 055 (1.00) Unit.1 has entered VEGP 19100-C, " Loss of All AC Power," because NO AC , emergency buses are energized. The following additional conditions exist:
-
Steam driven AFW pump will not start
-
All S/G levels are 10% (NR)
-
Reactor power is e 5% on all PR channels
-
Core exit TC's are 732 degrees F
-
Subcooling is O degrees F
- RVLIS is not functional At this point the crew should: -
a. exit to VEGP 19221-C, " Response to Inadequate Core Cooling."
b. exit to VEGP 19231-C, " Response to Loss of-Secondary Heat-Sink." , c. continue in VEGP 19100-C, " Loss of All AC Power."
_ _ implement VEGP 18038-1, " Operations from Remote Shutdown Panels."
QUESTION: 056 (1.00) With reactor power at 7%, which one of the following sets of conditions requires a reactor trip on Unit 17 a. PZR pressure 2285 psig; PZR level 18%; all SG levels 42% NR'.; ~ Train "A" SI has occurred, PZR pressure 2335'psig; PZR level 25%; all'SG levels 44%~NR; loss =of.13.8Kv bus 1NAA has caused the loss of-RCP's 1 and ' c. PZR pressure.2035 psig; PZR level 94%; all SG 1evels 40%;NR; steam pressure negative rate bistables are-tripped on three steam lines, d. PZR-pressure 1935 psig;;PZR level 90%; all SG levels 84%-NR; Turbine Trip signal.is_present.
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- _. . SENIOR REACTOR OPERAT ) Page 34 - QUESTION: 057 (1.00) Given the following plant conditions:
- Pressurizer pressure 985 psig - Pressurizer relief tank (PRT) pressure 5 psig PRT temperature 90 F -
Reactor is shutdown Assume ambient heat losses are negligible and the steam quality in the pressurizer bubble is 100%. Also assume pressurizer and PRT conditions do NOT chang Which one of the following PORV downstream temperatures would be caused - by a leaking pressurizer PORV? a. 230 F b. 260 F c. 300 F d. 340 F QUESTION: 058 (1.00) Refer to attached Critical Safety Function Status Tree (CSFST) F-0.4, Integrit Which one of the following describes the basis for limit A on Figure 1? - a. Gives the operator time to prevent a pressurized thermal shock condition, b. Indicates a potential for development of a flaw in the reactor pressure vessel if RCS_ temperature exceeds this limi c. Ensure the "cooldown rate" of the RCS is controlled to prevent permanent plastic deformation _of the reactor pressure vesse d. Prevents the growth of a flaw that could conservatively be: present in the vessel _ wal _ . . . . . .
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- SENIOR REACTOR OPERAT Page 35-
:
QUESTION: 059 (1.00) j Why are the RCPs required to be tripped at 1375 psig on a small-break-LOCA? i i a. To eliminate RCP heat input into the RC i b. To reduce inventory loss through the break, c. To prevent thermally shocking the reactor vesse d. To prevent RCP seal damage due to excessive seal injection flo L t F QUESTION: 060 (1.00) At some time after a large break LOCA, recirculation is transferred from cold leg recirculation to_ hot leg recirculation. The primary reason hot-leg recirculation is used at this point-rather than continuing to use-cold leg recirculation is: a. to conserve RWST inventor b. to prevent exceeding the boric acid solubility limit in'the , Cor c. to stabilize RCS temperatures when a pressurized thermal shock .
!
condition is imminen to collapse any voids that-have formed in the steam generator ,
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__ - . _ _ _ - _ _ - . . . _ . . _ . _ . . . _ _ . . _ _ _ - - _ - _ . _ _ _= _ . _ . _ ._.___ _ _ . . - - SENIOR REACTOR OPERAT ) Page 36 QUESTION: 061 (1.00) ] A steam generator (SG) tube rupture has occurred, and VEGP 19030-C, i
" Steam Generator Tube Rupture," is being execute Before the operator ,
'
can commence the RCS cooldown and depressurization per VEGP 19030-C, the ruptured SG pressure must be checked to be greater than 290 psi Why is the 290 psig limit imposed?
'
a. To ensure that the operator can block the low stcamline pressure SI signal, which would actuate below 290 psi b. To ensure RCS pressure will be less than ruptured SG pressure after the cooldown to stop primary-to-secondary leakage, c. To ensure a PTS condition for the reactor vessel is not developed during the cooldow > d. To preclude a return to criticality during the rapid RCS cooldow QUESTION: 062 (1.00) The following plant conditions exist:
- Unit 1 is at 25% powe RCP TRIP annunciator is li RCP LOOP 4 LOW FLOW ALERT annunciator is li RCP 4 MTR OVERLOAD annunciator is li Which one of the following actions is an IMMEDIATE ACTION for this event per VEGP 18005-C, " Partial Loss of Flow"? Check RCP 4 has tripped and attempt to restar b. Close Loop 4 spray valve-(PIC-0455B).
c. Verify pressurizer level'is trending to progra d. Verify No. 4 steam generator level __is trending ~to_65%.
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i SENIORREACTOROPERATd - Page'37 _ QUESTION: 063 (1.00) Which one of the following conditions requires emergency boration to be performed? a. Reactor Engineering reports that shutdown margin equals 1.3% delta-k/k while in MODE b. Intermediate range power (amps) has doubled in three minutes while in MODE c. Control rod bank D stepping out without operator action while in MODE _ d. SOURCE RANGE HIGH FLUX AT SHUT DOWN annunciator in alarm while in MODE QUESTION: 064 (1.00) With Unit 1 in MODE 1, the following annunciators have just alarmed:
- ACCW RCP 3 CLR LO FLOW annunciator - ACCW RCP 3 CLR OUTLET HI TEMP annunciator - ACCW RTN HDR FROM RCP LO FLOW annunciator According to'VEGP 18022-C, " Loss _of Auxiliary Component Cooling Water," -which one of the following conditions requires tripping No. 3 RCP?
a. Seal water outlet temperature at 200 degrees b. Motor stator winding temperature at-200 degrees c. Motor lower radial bearing temperature at'200 degrees Pump lower seal water bearing temperature at 200 degrees _ . _ _ _ _ _ _ _ _ _ . _ . . _ _ _ _ _ . _ _ _ _ _ . _ . _ _ _ _ _ _ _ _ __..._. _ - . SENIOR REACTOR OPERA Page 38
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QUESTION: 065 (1.00)
:
The following plant conditions exist:
, - Unit 1 is at 100 percent power with pressurizer pressure control in Aut Pressurizer pressure channel selector is in the. 455/458 positio Pressurizer pressure transmitter PT-458 fails to 2390 psi Which one of the following actions or conditions will_STOP further plant ;
degradation from this event? a. Pressure transmitter PT-456 senses less than 2150 psi b. Pressure transmitters PT-455 and PT-457 sense less than'2185 psi c. The operator manually closes PORV PCV-455 d. The operator nanually closes spray valves PCV-455B and PCV-455 ' QUESTION: 066 (1. 0 0 ) ' If rod control is in AUTO when symptoms of a possible narrow range temperature instrument failure are noted, what action should be-taken by the operator BEFORE the control rods are placed-in MANUAL 7 a. Verify the rods are stepping I b. Verify the temperature instrument is NOT in test, c. Verify turbine runback or load shed is.NOT in progres ,
,
d. Verify reactor trip should NOT have occurred due to the failure.
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SENIOR REACTOR-OPERATOt - Page 39 QUESTION: 067 (1.00)
.The following conditions exist: - The unit is in Mode RHR Train A is in operatio Reactor vessel level is at 188 feet (mid-loop).
i RHR Heat Exchanger Bypass Flow Controller (FIC-0618A) has just failed, causing RHR flow to increase to maximum. If no operator action is taken and RHR flow remains at maximum, which one of the following would_ occur to cause a loss of RHR? a. Pump overspeed trip from runout due to low discharge pressur b. Pump overcurrent trip due to the high discharge pressur c. Loss of pump suction due to gas entrainment in the loop suction lin d.-Loss of pump suction due tollow net positive-suction head.- QUESTION: 068 (1.00) Which one of the following RCS leak rates is within allowable limits per the plant Technical Specifications? a. 8 gpm identified leakage, gpm unidentified leakage, gpm through RHR suction valve (HV-8701A).
d. 8 gpm steam generator tube leakag j
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_ . . - _ _ _ _ _ _ . _ . _ _ _ _ _ . _ _ J t 8 i j SENIOR R2 ACTOR OPERA ' V Page 40
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i i
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! { QUESTION: 069 (1.00) ! Control rod H8 has just been withdrawn 15 steps to realign it with I control bank (CB) D in accordance with VEGP 18003-C, " Rod Control System i Malfunction." After the rod was realigned, the P/A converter for CB D l l was NOT reset to the original bank heigh i
i l- Which one of the following will occur because of this oversight? l-i
- a. ROD CONTROL NON URGENT FAILURE alarm when CB D is inserte l l !
' b. ROD CONTROL URGENT FAILURE alarm when CB D is withdrawn, j i
c. Spurious ROD DEV/ RADIAL TILT alarm ~ i l=
}
d. Improper ROD DANK LO LIMIT alarms.
l
i- i
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, QUESTION: 070 (1.00): ! I i , Which one of the following explains why it is important to shift rod l ' control to AUTO during a loss of main feed pump with plant power at 90%? j j a. To reduce Tavg to 557 degrees F, thereby increasing the DNB !
-
l b. To ensure Tavg is not reduced below the minimum temperature for criticalit i ! !
' c. To maintain Tavg accurately on program for the indicated steam- . flo !
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- To match Tavg to Tref-more quickly-than manual rod control ca '
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SENIOR REACTOR OPERATO Page 41 QUESTION: 071 (1.00) Which one of the following describes No. 1 steam generator (SG 1) response to a trip of No. I reactor coolant pump at 25% reactor power? a. SG 1 level will " shrink" and then increase due to FRV leakage, b. SG 1 will continue to steam at 90% of the pre-event rat c. SG 1 level will der le to the low level trip setpoin d. SG 1 pressure will . crease to saturation pressure for RCS Tav QUESTION: 072 (1.00) The following plant conditions exist:
- - Unit 1 has tripped from 100% due to a loss of off-site-powe Both diesels have started and the sequencers are loading the buse seconds after sequencer start, SI is inadvertently actuate Which one of the following explains the-subsequent response of the Unit i sequencers to the SI actuation signal?
a. The sequencers must be reset nanually before -the SI loads will begin sequencin b. The sequencers will reset immediately and start sequencing the SI loads, ' c. The sequencers will reset in 20 seconds and start-sequencing-the SI load d. After 20 seconds the sequencers will start sequencing SI loads not started previously by the'UV sequencer.
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SENIOR REACTOR OPERATO Page 42 QUESTION: 073 (1.00) Unit 1 is operating at 100% with Bank D rods at 218 steps. An electrical failure has deenergized the "C" 120V AC Vital Instrument Bus -
(1CY1A) and now you note that the rods cannot be withdrawn in auto or manual. Which one of the following is preventing outward rod motion?
a. C-1, IR overpower rod sto b. C 2, Power Range High Flux Rod Sto c. C-3, Overtemperature Delta-T Rod Sto C-4, Overpower Delta-T Rod Sto QUESTION: 074 (1.00) A small fire causes-the loss of bus 1AA02. The control room operators-- - - should perform which one of the following IMMEDIATE operator actions: a. Emergency stop A diesel generator, b. Commence a plant shutdown to hot standb c. Check A diesel generator running and attempt to energize 1AA02 after the fire is ou d. Attempt to close A diesel generator output breaker; if it won't close, normal stop A diesel generato QUESTION: 075 (1.00) Priority keys are used for access to: a. vital areas, b. radiation area c. locked high radiation areas, g- d. safeguards information area .
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SENIOR REACTOR OPERA 70 Page 43 l l
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QUESTION: 076 (1.00) i In accordance with VEGP 00150-C, " Deficiency Control," the USS should ! complete his review of a new Deficiency Card within after ; receiving i , i a. 2 hours b. 4 hours
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' hours d. one shift
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QUESTION: 077 (1.00) -
;
I A containment entry is to be made with the incore detectors not fully inserted but tagged ou In accordance with VEGP 00303-C, " Containment Entry," which one of the following statements applies? a. The USS can grant permission to enter' containment-only after h logs the status of the incore detectors, b. The USS can grant permission to enter containment only after the- -i HP supervisor evaluates the radiological consequence , c. No special considerations are required'before the USS can grant permission to enter, containment because the incore detectors are- , tagged.ou d. The-USS-can-grant permission te enter containment-after-the itP superintendent evaluates the-radiological 1 consequences and management-concurrence is_-obtaine .
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SENIOR REACTOR OPERATO Page 44 QUESTION: 078 (1.00) The Unit Shift Supervisor (U: ' has determined that a clearance must be released on a system, but a s- clearance has not been released,- If the subclearance holder (SCH) is not on site and cannot be contacted, the subclearance can be released provided: a. The foreman for the SCH gives the Shift Superintendent (SS) or his department supervisor permission by phon b. The foreman for the SCH gives the Shif t support Supervisor (SSS) permission by phone and will be responsible for notifying-the SCH and his cre c. The SS and department supervisor conduct a joint review / approval of the status of the subclearance work and-the need for an immediate clearance release, d. The SSS and the USS determine that the reason for the subclearance is no longer valid.
QUESTION: 079 (1.00) Which one of the following is considered a temporary modification, per the requirements of VEGP 00307-C, " Temporary Modifications?" a. A new style packing gland _ installed on a portable turbine-building sump pump under a work control program (to-be replaced after evaluation). ; b. A temporary jumper to defeat the self-testing circuit of the ESF Sequencer (to be removed after shutdown), c. Installation of_ leak repair sealant on a condensate system valve, where the pressure boundary is not breached, d.-A design change to the card reader system impacting the security pla SENIOR REACTOR OPERATO Page 45 i
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QUESTION: 080 (1.00) !
!
When Unit 1 is in Mode 5, and fuel movement is taking place on Unit 2, .i what is the minimum licensed operator shift crew manning requirement? l l a. One Shift Superintendent, one USS in each unit, one RO in each _ unit, and one SRO in containment-supervising fuel handling.-
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b. One Shift Superintendent to supervise fuel movement, one USS in each unit, and one RO in the control room for each unit,
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c. One RO, one Shift Superintendent, and one USS in the control
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room for each unit, and one SRO in containment supervising fuel ' handling.
- d. One SRO in containment supervising fuel handling, one_Dhift Superintendent, and one RO.in the control room for each uni , . QUESTION: 081 (1.00) . Filling in an LCO status sheet can be waived by the USS in which one of the following situations? a. LCO entry does NOT extend past the current shift, and appropriate USS 109 entries are mad b. LCO entry does NOT extend past the current shift, and Shift i Superintendent (SS) approval is obtained, c. LCO entry does NOT extend past 24 hours, and appropriate USS log entries are made, d. LCO entry does NOT extend past 24 hours, and SS approval is obtaine _ _
-
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_. - SENIOR REACTOR OPERATO Page 46-
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i QUESTION: 082 _ (1.00) ! Which one of the following actions is the specific responsibility of the , USS during Mode 1 operations upon receiving a report that an unmonitored ' manual suction valve for D train Safety Injection (SI) pump is jammed ! closed rather than being in its required locked-open position? l
,
n. Verify power is available to the Safety System Monitoring Panel .
(SSMP). !
b. Manually energize the SSMP light for the B train of the SI syste , c. Review the valve lineup for the SI system by ordering a walkdown, d. Check the Locked Valve Manipulation Log to discover _the-person responsibl .
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QUESTION: 083 (1.00)
-Which one of the following describes the purpose of the interlock which requires the CVCS positive displacement (PD) _ pump bypass valve to be OPEN when the pump is started?
a. To permit the pump-to start in an unloaded conditio , b. To prevent a pressure spike on the pump discharge piping, c. To provide sufficient flow to minimize heat buildup in the pum , d. To ensure adequate Net Positive Suction Head to the pump during [ startu [
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SENIOR REACTOR OPERATO Page 47 QUESTION: 084 (1.00) Which one of the following describes the resulting accuracy of the DRPI System if-a Data A failure occurs? a. + or - 4 steps b. + or - 10 steps , + 4 steps d. + 10, - 4 steps QUESTION: 085 (1.00) During operation at 50% power at Unit 1, Auctioneered HI Tavg to.the-pressurizer level control program fails to 600 degrees Which one o the following describes the response of the plant if no operator action is taken? a. Low Level Deviation alarm will actuate, charging flow will increase, actual level will stabill:e at 60%.. b. Low Level Deviation alarm will actuate, charging flow will increase, actual level.will rise causing a reactor tri c. High Level Deviation alarm will actuate, charging flow will decrease,' actual level will stabilize at 25%. d. High Level Deviation alarm will actuate, charging flow will1 decrease, actual level will.-lower to:17% securing. letdown.- -
- . - - - . . . . - - _ _ - ~ - . - - - . _ - - - . . - . - . . - - - - . - - l SENIOR REACTOR 10PERATOh O Page 48
i QUESTION: 086 (1.00) Refer to the attached Tech Specs Figure 2.1-1 for the Reactor Core Safety Limits. !&2tch each labeled 2250 psia curve segment in Column A with the appropriate basis for the segment from Column (0.5 each) ,
(Note: Numbers from Column B may be used once, more than once, or not at all, but only ONE answer may occupy each answer space.}
Column A Column B (CURVE SEGMENT) (BASIS) Segment I Prevents exceeding DNBR limit ' anywhere in cor i Segment II ' Prevents exceeding 15% quality in coolant at core exi . Prevents T-hot from reaching saturatio . Prevents exceeding fuel centerline temperature of 2200 degrees . QUESTION: 087 (1.00) i Which one of'the following is the basis for the 2200 degrees F limit on fuel rod cladding temperature? a. This temperature causes a total hydrogen' generation of-100 percent of the theoretical maximum hydroge b. This temperature causes excessive non-condensible gas accumulation in the RC c. This temperature causes accelerated cladding oxidation,_which can cause cladding failur d. This temperature causes the Zircaloy-water reaction to become exothermi .__- _ _ _ _ . - _.--.-_ _ _ - . . _ _ _ . _ _ . _ . . . _ -,_ a . . ,.=_ _ _ _ _ s
-
O O
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% '
OPERATION
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660 2440 pais 650 \ ' N : i
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h 630 N
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2000 psia % 1
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620 N y N 'N g u 3 g se cuenry U 610 ' ~
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C i 1935 csia T % } 600 i 7 ACCEPTABLE-OPERATION 590 580 .1 .2 .3 .4 .5 .6 .7 .8 .9 .1 * FRACTION OF RATED THERMAL POWER FIGURE 2.1-1 REACTOR CORE SAFETY LIMIT
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, V0GTLE UNITS - 1 & 2 2-2 Amendment No.44 (Unit 1) ) Amendment No.24 (Unit 2) j
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SENIOR REACTOR OPERAT Page 49 QUESTION: 008 (1.00) -In accordance with the foldout page of VEGP 19001-C, " Reactor Trip Response," which one of the following combinations of conditions indicates that a RED path condition exists for core cooling? ANY RCP CORE RVLIS LEVEL RUNNING EXIT TC (FULL RANGE) a. NO 700 F 30% b. YES 750 F 80% c. NO 800 F 35% d. YES 850 F 90% QUESTION: 089 (1.00) ,
.
After a reactor trip, Critical Safety Function Status Tree (CSFST)- monitoring must begin when the crew:
,
a. has completed the immediate actions of VEGP_19000-C, " Reactor Trip or Safety Injection."
b. enters the EOPs (beginning with step 1 of VEGP 19000-C). ;
;
c. has identified the symptoms of a challenge to a_ radiation barrier, no matter where in the EOPs the challenge occur , d. exits-from VEGP 19000-C--into an optimal recovery path (ORP) or when step 28 of 19000-C is reached, whichever-comes'firs , _ i
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-SENIOR REACTOR OPERAT ) ) Page 50-QUESTION: 090 (1.00) VEGP 19211-C, " Response to Nuclear Power Generation /ATWT," was implemented when the reactor would not trip upon receipt of a reactor trip signa The control rods are being manually inserted by the Reactor Operator. When reactor power is checked in Step 4 of VEGP 19211-C, which one of the following sets of conditions allows the operator to return to VEGP 19000-C, " Reactor Trip or Safety Injection"? CONTROL ROD IR STARTUP REACTOR MOVEMENT RATE (DPM) POWER (%) In .5 8 Stopped +.2 2 In 0 6 Stopped .3 3 QUESTION: 091 (1.00) Which one of the following conditions will cause the SMALLEST increase in area radiation monitor indications in the spent fuel pool area? - a. Lifting the core source element out of the poo b. A large tritium release in the new fuel area of the pool, c. A drop in pool level to the bottom of the transfer gate seal, d. Radiography of welds in the transfer canal are , a
SENIOR REACTOR OPERAT ) Page 51 QUES"JION : 092 (1.00) RCS pressure is increasing past 2335 psig due to loss of the secondary heat sin One pressurizer PORV is jammed shut, and the other presse;i'er PORV block valve will NOT open. According to VEGP 19231-C,
"Respt-~ e to Loss of Secondary Heat Sink," which one of the following actions should be taken in response to this situation?
a. Stop both CCP3 to limit the pressure transien b. Allow RCS pressure to increase until the pressurizer safeties lift to control pressur c. No action is required since this situation does not meet the - entry conditions for VEGP 19231- d. Open the reactor vessel head vent valves to establish a bleed 4 pat QUESTION. 093 (1.00) VEGP 19012-C, " Post LOCA Cooldown and Depressurization," is used following a LOCA event in which the RCS remains at pressure, and it is determined that: a. SI termination criteria cannot be met or maintaine b. SI termination criteria have been met and can be maintained, the possibility of core voiding needs to be prevente d. a pressurized thermal shock condition is imminen SENIOR REACTOR OPERAT - Page 52
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- QUESTIONi 094- -(1.00)
-
I
.During the transfer to hot leg _ recirculation, it is discovered _that the RER Train-B to Hot Leg Crossover Valve-(HV 8716B) is jammed shut and -
.
will NOT open. The proper course of action for this situation is to: a. shift train A to hot leg recirculation and shut deva rain'BL equipmen ; b. place train A in hot leg recirculation and maintain / realign train B in cold leg recirculation, c. maintain / realign both trains in-cold leg-recirculation, d. terminate ECCS' flow and realign charging to normal charging pat QUESTION: 095 (1.00) In VEGP 19030, " Steam Generator Tube Rupture," the RCP trip criteria NO longer apply if: a, the SI signal is rese b, the RCS cooldown is starte , c. the ruptured SG is isolated, d. executing any steps of VEGP 1903 ._ __
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' - SENIOR' REACTOR OPERAT -
Page 53-
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t QUESTION: 096c (1.00) ;
-The plant _has stabilized.following a major tube rupture in SG-1. :The following-conditions exist: - ECCS flow has been terminate Normal charging and letdown have been establishe Aux feed flow to the ruptured SG is isolate SG 1 level is 75% N The steam dumps are availabl In accordance with VEGP 19030, " Steam Generator Tube Rupture," which one-of the following is the preferred cooldown method for-SG 1 under these - conditions?
a. Dump SG 1 steam to atmosphere; feed SG 1 with AF b. Use all SGs including SG 1, dumping steam to the condenser;' feed-SGs with AFW, c. Reduce RCS1 pressure, allowing rupture backflow and maintain ~SG 11 ' level with AF d. Use SI flow into the-RCS; dump steam using-SG 1 AR QUESTION: 097 (1.00) The following plant conditions exist:
- Unit 1 has tripped from power due-to loss of instrument ai All attempts.to restore instrument air have faile Which one of the~following is the-directive.for entering MODE!4-found i VEGP_18028, " Loss of' Instrument Air"?
a. Cool down to MODE 4 rapidly to reduce plant stored energy.
' b. Cool down to MODE 4 rapidly'to reduce release rates through-
-fail-open--valves, c. Remain in MODE 3 to provide time to restore instrument ai d. Remain in MODE 3 to ensure that_VEGP 19001-C, " Reactor Trip .
Response," is. complete.
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LSENIOR REACTOR'OPERAT , Pag'e 54-EQUESTION 098. _(1. 0 0 ) - VEGP 93500-C, " Manual Operation of Fuel Handling Equipment," describes manual actions taken to accomplish which one of the-following? a. Manipulate new fuel during receipt inspection and transfer to-the-spent fuel pool, b. Continue refueling offload operations in the event of a containment building power failur c. Place fuel assembly in safe condition in the event'of a fuel handling component failur , d. Move RCCAs between fuel bundles in the spent fuel poo QUESTION: 099 (1.00) VEGP 19231, " Loss of Secondary Heat Sink," requires.immediateoinitiation Hof feed and bleed operations if no CCPs are available during a loss'of; heat sink event, no matter what the steam generator levels are. - What'is the reason for proceeding directly to feed and bleed operations in this situation? a. To prevent adding cold feedwater to afdry SG and possibly ? rupturing-the S .b. To ensure RCS pressure can be lowered enough to allow SI injection for feed and bleed operation c. Time considerations require initiation of feed and bleed.within 30 minutes, d. To preclude-SG-pressure from increasing.enough to stopffeed' flow from the condensate pump .
.
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- _- _ _ _ _ _ _ _ _
SENIOR REACTOR OPERAT Page 55 QUESTION: 100 (1.00) A LOCA in containment is causing a radioactive release from the plant ven Which one of the following methods will provide the most ACCURATE off-site dose projections for the Shift Supervisor prior to TSC activation? a. The default release rate from the VIBRANT computer code for a LOC Isotopic analysis of reactor coolant times the default containment vent flow rate under existing conditions, c. RE-12442 monitor reading times the plant vent flow rate from the ERF compute RE-0005/0006 monitoring reading times the default containment leak rate (0.2% volume / day).
(********** END OF EXAMINATION **********) ,
- - -
SENIORREACTOROPERATd Page 56 ANSWER: 001 (1.00) a.
REFERENCE: VEGP 00304-C, pg. 10 KA 194001K102 [3.7/4.1] 194001K102 ..(KA's)
-
ANSWER: 002 (1.00) REFERENCE: VEGP 00304-C, pg. 22 KA 19tiG01K102 [3.7/4.1] 194001K102 ..(KA's) ANSWER: 003 (1.00) - REFERENCE: LO-LP-63308-09-C, pg. 10 KA 194001K101 [3.6/3.7] 194001K101 ..(KA's)
.. ...=- . . . - - . . . . . - . - .- .-. - .. .- - . ..~ , . . . _ . - . . .-
SENIOR REACTOR OPERA Pag e 5 7 -_-- ,
.
-ANSWER:- 004 (1.00)
.
,
' ,
REFERENCE: VEGP 00404-C, pg. 6-KA 194001A103 [2.5/3.4]' 194001A103 ..(KA's)
ANSWER: 005 (1.00) REFEREN("E : VEGP :.0000-C, pg. 1 KA'194001A111 [2. 8/4.1) . 134001A111 ..(KA's) 't
- ANSWER: 006 (1.00) ' REFERENCE:
. VEGP 00920-C, pg. 3
=KA-194001K104 - [ 3.3/3.5]
, 194001K104 . ..(KA's)
.
t +
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~
SENIOR. REACTOR OPERA .
--
Page.58?
; ANSWER: -00 (1.00): REFERENCE: -VEGP 00308-C, Sect. 2.3, 4.1. ,
KA 194001K104 [3.3/3.5] 194001K104 ..(KA's) ANSWER: 008 (1.00) ' REFERENCE:
,
_VEGP 00920-C, pg. 7
,
KA 194001K103 [2.8/3.4] 194001K103 ..(KA's) ANSWER: 009 (1.00)
. REFERENCE:
VEGP 00920-C,: pg. 5 KA1194001K10 [2.8/3._4]
.194001K103 ..(KA's)
.
--wy ,- , -wmn
_ ._ SENIOR REACTOR OPERAT - PageL59- -ANSWER: 010 (1. 00 ): REFERENCE: LO-LP-27102-10, pg.. 13, L.O.-5 KA 001000Alf6 (4 .1/4 . 4 ] 001000A106 ..(KA's) _ ANSWER: 011 (1.00) REFERENCE: LO-LP-27012-10, pg. 16, L.O. 9 KA 001000K504 [4.3/4.7] 001000K504 ..(KA's) ANSWER: 012 (1.00) REFERENCE: LO-LP-16401-14, pg. 12 KA 003000K201 (3.1/3.1] 003000K201 ..(KA's)
- - - - -
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: SENIOR-REACTOR OPERAT .
_ Page 60'-
,
< i. ANSWER: .. 013 .l(1. 00 )
. =
- REFERENCE: _
.,
LO-LP-16401-14,;pg'. 11, L.O. - 8 KA 003000K201 (3.1/3.1] 003000K201 . . (KA's) c ANSWER: 014 (1.00) , , , REFERENCE: * , LO-LP-09101-08, p. 9, L.O. 8 KA 004010A204 (3.6/4.2] 004010A204 . . (KA's) ' , ' ANSWER: 015- (1.00) d.
. REFERENCE: LO-LP-09501-05, pg. 11, L.O. 6-
--KA;004000K104 (3.4/3.8]
F 004000K104 . . (KA's)
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'
..
'
SENIOR-REACTOR.OPERAT V - : Page 61-
,
i
:
ANSWER:- 01 (1. 0 0') -
- . i - REFERENCE: --VEGP 13009, Section 4.6~
KA 004000A207 (3.8/3,9)- , 004000A207 ..(KA's) ANSWER: 017 (1.00) REFERENCE: ,
-LO-LP-28103-13-C, pg. 32, KA.013000K103 [3.8/4.1)
013000K103 ..(KAs) ANSWER: 018 (1.00) ,
- ' REFERENCE;. "
LO~-LP-28101-10-C, pg. 13, L.O. 6 KA-013000A301 (3.7/3.9) 013000A30 ..(KA's)'
.
-
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. _ - -. - -- . . . . . . . . . - . - - -
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SENIOR-REACTOR OPERAT - Page'62
,
;
- ANSWER: 019' (1.00) ?
- i
. REFERENCE: >
-LO-LP-36102-04-C, pg. 14, L.O. 7 KA 017020A201 - [3.1/3.'5]
.
017020A201' ..(KA's)- :
:
ANSWER: 020 (1.00)
- REFERENCE:
.
'-'
-LO-LP-29130-04, pg.-6, L.O. 2
,
.KA 022000A301 [4.1/4.3]
.
022000A301 ..(KA's) , ANSWER: 021 (1.00) d-. --
.. .. REFERENCE:
F LO-LP-18201-13-C, pg. 13, L.O. 7
l- KA-059000K419 [ 3 . 2 / 3 .' 4.]
'059000K419'
-
..(KA's)-
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SENIOR REACTOR OPERAT -'Page 63 . ,
,
ANSWER: 022' ( 1. 0 0 )' L REFERENCE:
: -LO-LP-20101-17 C, pg.-11, KA 061000K404 (3.1/3.4]
061000K404 ..(KA's) ANSWER: 023 (1.00) REFERENCE: LO LP-20101-17-C, pg. 12, , KA 061000K103 [3.5/3.9) 061000K103 ..(KA's) ANSWER: 024 (1.00) REFERENCE:
.LO-LP-20101-17-C, pg. 12, L.O. 4 KA 061000A102- {3.3/3.6]
061000A102- ..(KA's)
- . , , . _ _ ._. ._ _ .--. __ . . _ _
. __
. . . . .
- . - SENIOR REACTOR OPERAT ; Page 64 TANSWER: 025' (1. 00 )-' REFERENCE:
LO-LP-01001-04, pg. 11, L. KA 062000A401 (3.3/3.1]
-062000A401 ..(KA's)
._
ANSWER: 026 (1.00) C.-
-REFERENCE:
LO-LP-16303-12-C, pg. 8, KA 010000A403 (4.0/3.8] 010000A403 ..(KA's)
' ANSWER: 027 (1.00) - REFERENCE:
T LO-LP-13001-11-C, p , L.O. 2-KA 006000G004 ( 3 . 5 / 3 . 8 ]--
-006000G004 ..(KA's)
.
un irassir Tsiu m i - i
. .
..._.- ,.. . . _ . _ . . . . ,. . _ . _ _ _ _. .- . . . . - , . . , _ _ _
-
. . . . .
SENICR-REACTOR OPERAT - Page ~. 6 5 - - ANSWER: 028-- ' (1= . 0 0 ) a.
'
- - REFERENCE:
LO-LP-16001-11, pg. 23-24, L.O. 1-KA 002000K106 (3.7/4.0} 002000K106 .4 (KA's) ,
ANSWER: 029 (1.00) ,
- REFERENCE:
LO-LP-16701-08-C, L.O. 5 KA 002000K402 [3.5/3.8] 002000K402 . . (KA's) ANSYdR: 030 (1.00)- a REFERENCE: LO-LP-25702, ,pg. 20; VEGP ARP 17005-KA-033000K303 - [3. 0/3. 3] . .. st
-- 03 3 00 0K3 03 - . .- (KA's)'
. . -. - . . - .
- _ _ _ _ __ _ _ _ _ _ _ _--_ __-__
SENIOR REACTOR OPERAT Page 66 ANSWER: 031 (1.00) REFERENCE: LO-LP-17101-07-C, pg. 14, KA 015000K301 [3.9/4,3] 015000K301 ..(KA's) ANSWER: 032 (1.00) C, PEFERENCE: ' LO-LP-17201-08 C, pg. 11, KA 015000A303 [3.9/3.9] 015000A303 ..(KA's) ANSWER: 033 (1.00) REFERENCE: LO-LP-28103-13-C, pg. 23, KA 013000A402 [4.3/4.4] 013000A402 ..(KA's)
,
. .
.+ . . ~ _- . - , . . .... - . ~ . .
, . . . .. -. _.- -.. ... - . .
"
- SENIOR REACTOR OPERA ~Page-67)
,
'
ANSWER: 034 --(1 00)
- ,
'
.-
)
- REFERENCE:
~ LO-LP-28101-10-C, pg. 12, . KA 012000K401 [3.7/4.0] 012000K401 ..(KA's)-
.
ANSWER: 035 (1.00) REFERENCE: , LO-LP-20102-08, pg. 8, L.O. 2 KA 012000G008 [3.9/3.8]. 012000G008- ..(KA's)
- ANSWER: '036 . (1. 0 0 ) ' -
C '. . REFERENCE: LO-LP 15101-10, pg. 15 KA 027000A401- [3.3/3.3]
'
027000A401 ..-(KA's) ,
..
g g g u- ver- '
--
. SENIOR REACTOR OPERAT -
Page 68: ANSWER:- -0371 -(1.00)
- REFERENCE:
LO-LP-29140-07, p , L.O. - 2
.}Gk 028000G005 [3.0/3.6) -028000G005 . (KA's)
_ ANSWER: 038 (1.00) REFERENCE: LO-LP-11201-14-C, pg. 55, LO-15 KA 064000G007 .(3.4/3.6) 064000G007 ..(KA's) ANSWER: 039 (1.00) - : REFERENCE:
.LO-LP-11104-0-C, pg. 17, .LO-9-KA'064000A101 (3.0/3.1)
064000A101 ..(KA's).
.,
_ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ . _ . _ _ . _ . _ _ _ _ _ _ _ _ _ . _ ,
_......... SENIOR' REACTOR OPERAT Page.69 ANSWER: 040 (1.00) REFERENCE: LO LP-11202-15-C, pg. 60, L.O. 13 KA 064000A406' [3.9/3.9)
.
064000A406 ..(KA's) _ hNSWER: 041 (1.00) . REFERENCE: LO-LP-12101-30-C, pg. 11, KA 005000K201 (3.0/3.2] 005000K201 ..(KA's) ANSWER; 042 (1.00) - REFERENCE: LO-LP-10101-11, pg. 16, L.O. 5 KA.008030A304 {3.6/3.7) 008030A304 ..(KA's)
-
- .-_ _ -. - - - - - -___ ._ _ ._ - - -
-
SENIOR' REACTOR. OPERA . Page-[70- !.
.>
i; ir !; i ANSWER: 043 (1.00) * l:.
' - REFERENCE:
LO-LP-10101-11, pg; 10, L.O.-4 t KA:008000A301 [3.2/3.0} ! ! ij 008000A301 ..(KA's) ! ! !- h ANSWER: 044 (1.00) I I-li i l REFERENCE: l-- (' LO-LP-16301-12-C,-p. 24, 25, L.O. 13
-
,
I KA'007000K401 (2.6/2.9) , i: I 007000K401 ..(KA's) i i ! ! ANSWER: 045 (1.00) ' b.
jf L :-REFERENCE: .. _,
l = Steam Tables !. ' KA 010000G013' [3.5/3.7) [ q i: 010000G013 ..(KA's) r ! .;
.
E -l ANSWER: - 046 (1 00) { r , l- - .f < !- , t i e : L :s {. -I
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' SENIOR REACTOR OPERA .
Page'71-REFERENC VEGP Tech Spec'3.9.8'.2 VEGP 18020-C,;rev 5,. pg. 2 KA 008000K301 (3.4/3.5) 008000K301- ..(KA's)
-ANSWER: 047 (1.00) REFERENCE:
Ops Trng MCL CL-36 1.06
.KA 000001G003 [3.3/3.9)
A 000001G003 ..(KA's) ANSWER: 048 (1.00) REFERENCE: LO-LP-37002-09-C, pg. 6,7 KA 000011G010- [4.5/4.5) , . 000011G010 ..(KA's) ANSWER: 049 -(1.00) C.
'
,
f
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..- . -.- .. .~. .-..-. . . - - . . . . . . . . . . . . - . . , . .
SENIOR-: REACTOR' OPERA . Page-72
. " REFERENCE:
' LO-LP-37002-09-C, pg. '6 KA 000069G012 [3.5/3.5] , 000069G012- ..(KA's)
.
ANSWER: 050- (1.00) REFERENCE: LO-LP-37002-09-C, pg. 6 KA 000009G012 [4.0/4.1) 000009G012 ..(KA's) ANSWER: 051 - (1.00) , REFERENCE: VEGP__19011-C,-pg. 2 KA 000056A203 (3.8/3.9) 000056A203 ..(KA's) ANSWER: 052 (1. 0 0 )-- a.
i.-
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g g % v % w r - ' - - - - - - - - - - - -
. . . . . . - - . . . . - . . - . . - . . - . --.~. . . . . - . _ . . - - . . . ~ . _ - - . . . . ~ - , _ , - . - . , - . . . .
.. .. . >
Page 73
' - SENIOR: REACTOR OPERA i
LREFERENCE: -
.- - : ..
LO - LP -' 3 7 0 03 - 0 7 - C , . pg . 12 KA1000074A204' '(2.5/2.5] '
.
!
000074K208 ..(KA's)
'
.a
-}
I ANSWER: 053- (1.00) ; REFERENCE: , I LO-LP-37031-09-C, pg. 7, VEGP 19100-C, pg. 19
-KA 000055K302 - (4. 3 /4. 6]
000055K302 ..(KA's) , ANSWER: 054 (1. 0 0 )-
,- C .
- -
REFERENCE:
- -
- -VEGP_19000-C,.pg. 2 KA 000007A202 (4.3/4.6]
. , " 000007A202 ..(KA's) , $ ANSWER: 055 (1.00)
.
. ?.
V
3-8 e
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.. . . - - . , . . . .- - - . . . - - - . - - . . - . . - . . . . .
- SENIOR REACTOR OPERAT _ .- Page.74-REFERENCE:-
{
LLO-LP-37031-09-C, pg. 6 KA 000055G012 [3.9/4.0) .
,
000055G012 ..(KA's) . ANSWER: 056 (1.00) , REFERENCE: VEGP 19000-C, Att.-A KA 000029A209 [4.4/4.5] 000029A209 ..(KA's) ANSWER: 057 (1.00) .
-REFERENCE: . Steam Tables-KA 000008K302 [3. 6/4 .1]
000008K302 .,(KA's) ANSWER: 058 (1.00) d.
,
- - -. . . _ - - _ , - - , , - - _ .. .. .-
_ _ _ _ . ._.m._.___.__________.._ - _ . _ _ . . . _ . _ _ _ _ . _ . _ _ _ _ _ _ . _ _ _ . . _ _ . _ . . _ _ _ . . - . . f Page 75
'
SENIORIREACTO'R-OPERAT .
.
.
<
.
s ,
- REFERENCE:
l 'LOLLP-37071-03,.pg.!8
-- .
.i
- '
KA 000040K101 . [4 .1/4 . 4 ] b , 000040K101 ..(KA's) t-
. ANSWER: 059 (1.00)
!
l b.
REFERENCE: > '
.
LO-LP-37111-09-C, pg. 10
' KA 000009K323 [4.2/4.3] I
'
l 000009K323 ..(KA's)
'
I i -ANSWER: 060 (1.00) -
,
i i F b.
, t REFERENCE: !! LO-LP-37114-08, pg. 6
-KA 000011K313 - [3. 8/4. 2] > K313 ..(KA's)
t F I- ~' ANSWER: ~ 061 (1. 0 0) !,
- C.
,.. !'s
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SENIOR REACTOR OPERATd I ) Page 76 REFERENCE: Ops Trng MCL CL-37 4.17 KA 000038K306 [4.2/4.5) 000038K306 ..(KA's) ANSWER: 062 (1.00) REFERENCE: VEGP 18005-C, pg. 2 KA 000015G010 (3.4/3.4) 000015G010 ..(KA's) ANSWER: 063 (1.00) ~ REFERENCE: _ VEGP 17010-1, pg. 19 - KA 000024K301 [4 .1/4 . 4 ) 000024K301 ..(KA's) ANSWER: 064 (1.00) . i
SENIOR REACTOR-.OPERAT -- -
.
Page 77-REFERENCE:: 1.VEGP-18022-C,=pg. 4-KA 000026K303 '[4 0/4.2]-
.
000026K303 ..(KA's)
- ANSWER: 065 (1.00) j q
REFERENCE: VEGP Logics, Fig. 7.2.1-1, shts. 11, 18, & 19 * VEGP LO-LP-60301-07-C, . pg. 14 KA 010000K403 (3.8/4.1] 010000K403 ..(KA's) ANSWER: 066 (1.00) REFERENCE: LO-LP-60301-07-C, pg. 12 KA 000007A103 (4.2/4.1] 000007A103 ..(KA's)
. ANSWER: 067 (1.00)
C.
4
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. _ . . _ _ , . ~ . . . . _ _ -
- - - _ _ . _ - - .. _ _ _ _-- . _ _ _ . . _ . .. ______
.
( -.
SENIORLREACTOR..OPERAT
.
-
1 =Page 78'- l; REFERENCE: :- >
- -
VEGP'LO-LP-12101-30-C, pg. 35' _i-VEGP 13011-1, p g ._ 2'-
'
- KA 000025K101 (3.9/4.3)
L !' -00002SK101 ..(KA's) - '
.
e ANSWER: 068 (1.00) :i '. ,
,
REFERENCE: TS 3.4.6.2 l KA 000009K320 [3.5/4.3) . 000009K320 ..(KA's) ..
,
ANSWER: 069 (1.00)
j d.
, !' -REFERENCE: i -- ! LO-LP-60303-13-C, pg. 30
.
L i KA'000003K309 [3.0/3.5] ' l- 000003K309 ..(KA's)- i-i 070 (1.00)
'
._ ANSWER: ! [ d.-
! .- l , i
,_ n2 . . _ _ _ , _ _ , - . _ . _ - ~ _ _ _ . . _ - . . . _ . . _ . _ . _ - . _ _ _
. _ _ . . _ . ._ . _ _ _ ., _ _.__.- _ _ _ - _ =.. _ .__ _ . _._ ._-_-._. _ _ _ ..._ _ -
._ ._ -.. .-- .
.
SENIOR REACTOR OPERA Page 79 i REFERENCE:- !
!
VEGP LO LP 60314-06-C, pg. 6 i i KA 000054K304 [4.4/4.6) l 000054K304 ..(KA's) E MiSWER: 071 (1.00)
, L i
'
REFERENCE: LO-LP-60305-04, pg. 6 J KA 000015K102 [3.7/4.1) { i 000015K102 ..(KA's) ,
;
i ANSWER: 072 (1.00) , ,! REFERENCE: * VEPG LO-LP-28201-15-C, pg. 8 i (3.8/3.9) ' KA 000056A247 t
!
000056A247 ..(KA's)
'
,
t' M4SWER: 073 (1. 0 0 ) . , i
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-
- , . . . . . - - . . , . _ . . . . . . , ...,,,,.,*
_ _ - - _ _ _ __ _ _ . - _ - _ _ _ _ _ _ _ . _ _ . SENIOR REACTOR OPERN Page 80 r i REFERENCE: j LO-LP-60324-01-02, pg. 20 ; i KA 000057A217- [3.1/3.4) ;
:
i 000057A217 ..(KA's) ' i
!
!
ANSWER: 074 (1.00) l i' REFERENCE: 7 LO-LP-60323-02, pg. 5 KA 000067G010 [3.3/3.7) ,
;
000067G010 ..(KA's) i ANSWER: 075 (1.00) , REFERENCE: LO LP-63008-07-C-08, pg. 6 j KA 194001K105 -[3.1/3.4) l
:
194001K105 ..(KA's) ANSWER: 076 (1.00) '! i ; _ _ _ _ _ _
,.
.I
.9 k
iTJ m - . ~.,,:2:.,...,~e -- ..J,.--mw,._,. - .
-, . ,,, . . . ~ . . , . . . ~ . ~ . . - , . , , , , , , . , . . , , - - . . - + - . - . . . - . . , ,
_ . - . _ . - _ _ . _ _ _ . . _. .._ _ - _ _ _ _ _ . _ . . _ - . . . . _ _ _ _ . _ _ - _ _ _ - - - . _ _ . - . _ , SENIOR REACTOR OPERA Page 81-i
:
REFERENCE: -! I LO-LP 63150-08, pg. 12 !, KA 194001A106 [3.4/3.4] j
:
.
194001A106 ..(YN s)
!
:
ANSWER: 077 (1.00) l - t i REFERENCE: > t LO-LP-63303-03, pg. G
,
KA 194001K114 [3.6/3.3)
'
.
194001K114 ..(KA's)
!
ANSWER: - 078 (1.00) { t
;,
REFERENCE . e VEGP 00304-C, pg. 22 , KA 194001K102 (3.7/4.1) , 194001K102 ..(KA's) 1 i ANSWER: 079 (1.00) ; b.
, f C
,
y . -+.v,-- - - - .yr -- ew:=,+.E,-se-ww.---em,rs.--.-r--,---w.-r,ww,-w-- r .-%.,*-..--m =--- re E - w -r c .--~r -,- - -.m-~. .-.*----.---Er--m'--.,-,---e---
. . _ _ . _ _ . . . . . . _ _ . . . . . _ . _ _ _ . . _ _ _ _ _ _ _ . . _ _ _ _ _ .._ _ __ ._ _ _ _ _ _ _ .-_ ..__._ . . . . _ . t SENIOR REACTOR OPERAT Page 82' !
! -REFERENCE:
i LO-LP-63307-08, pg. 5, 6 KA 194001A103 (2.5/3.4]
;
#
194001A103 ' u)
.
ANSWER: 080 (1.00) ; .
,
REFERENCE: ! VEGP 10003-C, pg. 4 KA 194001A103 (2.5/3.4) , 194001A103 ..(KA's)
>
!
ANSWER: 081 (1.00) REFERENCE:
VEGP 10008-C, pg. 2 KA 194001A106 (3.4/3.4] , 194001A106 ..(KA's)
ANSWER: .082 (1.00) b, i
!
l l o
, -_,--,-,,._.,_--...,.,rm.--.,,. .. y,.,,,,,,.--,,,_..,-,_,.,..J.,o,~,.y . , , , - . - , , -
_ .-. ._ m._.__ _ . . _ . . _ . -_._. ._.. . . _ _ . ._ _ _ - . _ _ _ . . . . ._ . . . _ _ _ _ _ . . . a SENIOR REACTOR OPERA . Page 83 ; i REFERENCE: !
!
LO-LP-63505-07, pg. 7 l KA 194001A102 [4.1/3.9) ,
.
194001A102 ..(KA's) l r ANSWER: 083 (1.00)
:
I ; REFERENCE: LO-LP 09201-09, Rev. 9, pg. 20 KA 004000K404 [3.2/3.1) 004000K404 ..(KA's) ANSWER: 084 (1.00) REFERENCE: LO-LP-27201-08, L. KA 014000A102 [3.2/3.6] 014000A102 ..(KA's) ANSWER: 085. - (1.00) a.
' , -
e -,- m+~ - , 4, ww w ,e e am-,,, - , . . . , m .,,.m.y,-
r SENIOR REACTOR OPEkAT Page 84-
;
-REFERENCE: ,
,
LO LP-16302-08-C, pg. 15, ! KA 011000A104 [ 3.1/ 3 . 3 ) > t
!
011000A104 ..(KA's) j i
-
,
ANSWER: 086 (1.00)
, ' REFERENCE: ;
LO LP 39203-06, pg. 6
.
YA 000001G004 [2.8/3.8) 000001G004 ..(Ya's) ANSWER: 087 (1.00) !
!
REFERENCE: LO-LP-36101-06-C, pg. 8, L.O. 1 -i LO-LP-36001-09 C, pg. 22, L.O. 10 KA 000011K305 [4. 0/4 1)__ , t 000011K305 ..(YA's) ANSWER: 088 (1.00) I
;
- '
. . - . - ..4-.. -m--. -..---.-..-...,-m--em,,,, _m-..,e,.--e.e.,- or,w .- . . ,w .y -,7-.-.m . bw e y. , , - -,syvy,-3#-,-- ,,- ,,r,,- , ,,..-r., ,-..c
= _ _ _ = _ _ _ _ _ _ _ - _ _ - _ _ _. __ _ -
_ _ _ -______ _____ _ _ . _ __.__ _ _ _ . i h l SENIOR REACTOR OPERA - Page 85
)
>
l l REFERENCE: li i j VEGP 19001 C, pg. 18 j i- ! - KA 000074A207 [4.2/4.7) l.
! ! l 000074A207 ..(KA's)
a '
!
l ! l ANSWER: 089 (1.00) I
d.
I l
..
f - REFERENCE: _ l l i l LO.LP-37002-09 C, pg. 12 l
!
! KA 0000G9G012 (3.5/3.5) { l
l'
000069G012 ..(KA's) , ! ! ,
'
l ANSWER: 090 (1.00) ! ! l ! { : REFERENCE: l ' j i ' VEGP 19211-C, pg. 7 KA 000029G010 [4.5/4.5] -!
;
;
000029G010 ..(FA's)
,
'
. ANSWER: 091 (1.00)
.i
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)
o )
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_
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. _ _ . _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ ___ _ _ _ _
,
t l . SENIOR REACTOR OPERA Page 86 i
'
!
!
b REFERENCE: ( i , Basic Rad Con /HP !
!
L i -KA 000061K101 [2.5/2.9) l
!
l 000061K101 ..(KA's) l
;
- ;
, i a ANSWER: 092 (1.00) f ! , ! i < > i ! ! !
>
l REFERENCE: : , , , , i- VEGP 19231-C, pg. 14 ! * l KA 000054G012 [3.2/3.2] t
.
3 i ., l 000054G012 ..(KA's) i
.
t i y . ANSWER: 093 (1.00) ! i .; , a.
, l i ! . REFERENCE: I i l LO-LP-37112-07, pg. 6 ; J- f r
- {
KA 000009K321 [4.2/4.5] .ff k +
, 000009K321 ..(KA's) .{
- ANSWER
- _ 094 (1. _0 0 )
,
b.
1 ,.- , i I l.
, ! r
$
l
- - , , , - . . ~ _ _ _ - ,-.,, -- .a..-..,...-.~.-.----..-..--.:-..--..-----
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. . . _ _ _ . _ . . m_ _ __ _ _ - . _ . . _ _ _ _ . . _ _ . _ . _ . _ . _ . . _ _ _ . . _ . . _ . . . . _ _ _ _ _ . _ _ .
_ SENIOR REACTOR OPERA . Page 87- [
&
REFERENCE:
,
VEGP 19014-C, pg. 2 { Ops Trng MCL CL-37 4.13 KA 000011A111 [4.2/4.2)
'
.
000011A111 ..(KA's) ANSWER: 095 (1.00) ,
,
REFERENCE: Ops Trng MCL CL-37 4.19 ; KA 000038G012 (3.8/4.0) 000038G012 ..(KA's) ANSWER: 096 (1.00) REFERENCE: LO LP-37312-08-C, pg. 10 KA 000038A139 (3.6/3.7) . 000038A139 ..(KA's)
.
ANSWER: 097 (1.00) c.
, I I f_.
l l l 1 :=
~
!.
. - _ . ._i.,_,....~.,._m.,__,,,_m .-.,.,.-~,y- . . . - - . . . - . , , . . . . , . . - . ~ . . .
'
,...,~..l....,,,,. ,,i,,- , - , . , ,,,,~. ~4.,_--.
._m- _ _ _ _ . . _ _ . _ . _ . _ _ _ . . ~ . . _ _ . _ . . - ~ . - _ - - -_. _ _ _ . . . - _ _ _ _ . . _ _ . . _ _ . _
SENIOR REACTOR OPERA Page 88 i REFERENCE: ? VEGP-18028-C, pg.-11 KA 00006SK308 [3.7/3.9) ! i 000065K308 ..(KA's) ANSWER: 098 (1.00) 1 i REFERENCE: LO-LP 25301-06, pg. 26 , KA 000036A104 (3.1/3.7) 000036A104 ..(KA's) ANSWER: 099 (1.00) . REFERENCE: * LO-LP-37051-10-C, pg. 11 KA 000074K305 (4.2/4.5) 000074K305 ..(KA's) . ANSWER: 100 (1.00)
-C.
..
-
I i. a .. .:_:.,.....-_..a........_......;.~.. ,.,__.J....;.._ , . , , - - - , . _ . , - . , ~ , , , . - , . . . , ,._,-.,;.-__.....;.,,... ..
_ . . . . SENIOR REACTOR OPERA Page 89 b REFERENCE: VEGP 91304-C, pg. 4 KA 000060K104 (2.5/3.7) 000060K104 ..(KA's) _ J a
(********** END OF EXAMINATION **'********)
'
.
_ . _ _ _ _ _ . . . _ . - - - - - - - - - -- ----
... . _ - --
- _ _ _ _ _ . _
i SENIOR' REACTOR OPERA Page 1
' ANSWER KEY I l f
,
'
'
I !
!
t 1 MULTIPLE CHOICE 023 a
'
i 001 024
'
a c !
!
002 d 025 c ; ' t 003 a 026 e i
*
, 004 b 027 d i
:
i
^
005 b 028 a 006 d 029 a ;
. -
:
007 b 030 a i 008 c 031 a ! r 009 d 032 c i
010 d 033 a ,
+
i 011 b 034 b ; i , 012 b 035 'b' 013 a 036 c , 014 d 037 b - 015 d- 038 d -' I-016 b 039 a 017 -d 040- d 018 c 041 d - I' ,- 019 d 942 b 4 i l ,. 020 -a 04 , 021 d 044 a 022- d -- 045 b ;
.
b
?
'!
f -- - C,.J,-.,,,,....-..-~m,r,,,,,.w,...44-4.,,--
~ ....._,,-
,,,,--...-,.,-.--..- ..-~ . --
.e- ._~,- -. . . . ,
'
_ __ i SENIOR REACTOR OPERAT Page -2
;
'
A N.S W E R KEY l,
!
!
',
I- l d
'
; 046 c 069
;
I 047 b 070 d i , :
- i.
! 048 b 071 a ;
- - !
! 049 c 072 b l . i ' a 050 a 073 b
.
q r 1 051-. b 074 a i 1 -t !- 052 a 075 a i 053 a 076 a- t
.
i
- 054 c 077 d !
>
i ! 055 c 078 c ; I j 056 a 079 b !
,
057 c 080 d ,' i j j 058 d 081 a l 059 b 082 b c :
'
060 b - 083 a- j ,- 061 c 084 d * 062 d .' 085 a 063 d- 086 MATCHING l . .
!
064-- c a 3 . !- 065 b b -- 1
- .
- .066 c MULTIPLE CHOICE-067 c 087 c Ii
- i
). '068' .a- '088 c ! ._ , .i
+
_
-
_ . ._ L... ; - _ _:
...... . _. ___ _ _ . -_ _ _ . __ _ - _ _ . _ _ . . . ~ . _ . . . _ _ _ . - _ _ _ _ . . _ _ . . _ - - . _ _ ._
SENIOR REACTOR OPERAT Page 3
!
ANSWER KEY
.
,
089 d 090 d , 091 b
+
092 d 093 a , 094 b 095 b 096 c 097 c
,
098 c 099 b 100 c
,
t F
<
-_
I.
l-i (********** END OF EXAMINATION **********)- . . . . .,;.4..._ _. ; .. ., . . _ - . . _ - _ . _ . ~ .
. __. ,._.._.2._..-.__ - __ __ . _ -
- . .. _ ..- - .. .-. _ . -- . . . . - - - . . - . - - . . . . . --- . - . - .-... .- .- () TEST CROSS REFERENCN ) Page 1-SRO Exam PWR Reactor !
Organized by 0uestion Number {
<
l OUESTION VALUE REFERENCE -i t 001 1.00 9000383 ! 002 1.00 9000384 003 1.00 9000386 ^ 004 1.00 -9000387 005 1.00 9000388 006 1.00 9000394 ci 007 1.00 9000395 008 1.00 9000396 009 1.00 9000398 010 1.00 9000401 , 011 -1.00 9000402 012 1.00 9000403 ' 013 1.00 9000405 014 1.00 9000406
'
015 1.00 9000409 016 1.00 9000410 ' 017- 1.00 9000411 018 1.00 9000412 019 1.00 9000413 020 1.00 9000414 021 1.00 9000415 022 1.00 9000416 ., 023 1.00 9000417
-
.
024 1.00 9000418 ,
'
025 1.00 9000419 026 1.00 9000422 027 1.00- 9000424 028 1.00 9000425 029 1.00- 9000426 030 1.00 9000428 031 1.00 9000430 032 1.00- 9000433 033 -1.00 9000434 - 034 1.00 9000436 035 1.00 9000437-036 1.00 9000438 037 1.00 9000439 038 1.00 9000440 039 1.00 -9000441 ' 040 1.00- 9000442-041 1.00 9000444 042- 1.00 9000445 043 1.00 9000446-044 1.00 9000447 045 1.00 9000448 046 1.00 .9000449
=047 1.00 9000451 048 1.00 9000452 049 1.00 9000454
^-
'W-e-h ----wwwf aMif'9 prb'4 '
d w' -my',ad ly'4-V -* V w y,M:---Aw4d- +t1W-e s"' 9 9 e 7- #' WF191'#ig'p w-r g v.m' vSe g--gr--$h
. - _ _
_ . . _ . _ . _ _ . -_ . _ _ _ _ . . . . _ _ _ _ . _ _ . _ _ . _ . . _ . _ . . . _ . . , _ _ _ . . . _ . - . _
,
i
) TEST CROSS REFERENCd - Page. 2 t
SRO Exam PWR R e a c't o r
Orga n i z-e d by Queat i on N-u m b o r i i i < f QUESTION VALUE REFERENCE ?
!
050 1.00 9000455 051 1.00 9000457 l 052 1.00 9000459 053 1.00 9000460 054 1.00 9000461-055 1.00 9000462 056 1.00 9000464 057 1.00 9000465 058 1.00 9000470 059 1.00 9000471 , 060 1.00 9000474 061 1.00 9000476 , 062 1.00 9000478 063 1.00 9000479 * 064 1.00 9000482 065 1.00 9000485' 066 1.00 9000486 067 1.00 9000488 [ 068 1.00 9000490 . . 069 1.00 9000491 ! 070 1.00 9000493 071 1.00 9000495 - 072 1.00 9000499 , 073 1.00 9000502 074 1.00 9000503 075 1.00 9000378 076 1.00 9000380 077 1.00 9000381 , 078' 1.00 9000382 079 1.00 - 90003A5 080 1.00 9000389 081 1.00 9000391 > 082 1.00 9000397 083 1.00 9000408 084 1.00 9000431 085 1.00 9000435 086 1.00 9000450 087 1.00 - 9000453-088 1.00 . 9000456' 08 ,00J 9000458 090 1.00- 9000463 091- 1.00 - 9000466 092 1.00- . 9000468 093 1.00 9000473 094- 1.00 9000475 095 1.00 9000477 096 1.00 --9000481 097 1.00- 9000492 098 1.00 : 9000496-T
,n evy,--~~~-,vr6----w, w
-
+---m..we--- - - - , , ---r-.,w- .,%m.,.~# . -4-,,. -p.,,.--.v, r -#y-y , ~ r--.. . --,w,--..
. ..
_ _ _ _ _ _ _ _ _ _ . -_ __ ._ ___ _ _ _ _ _ _ _ - - - _ _ -
-- ___ !
i TEST CROSS REFERENCL Page 3-4 .. l H l SRO Exam PWR Reactor O'r-9 a-n i z e d
-
by Quest ion Number ,
,
i 1' s a i
5
1 OUESTI_QH VALUE REFERENCE s
099 9000497
'
1.00 " 100 1.00 9000498
...... ,
l 100.00 J
:
i
, ...... ! , .
100.00 : l
.
>
.
!, -
,
-
;
.[
l
i l .i
;
k I 1 ..
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i ! i . . k I h !' 2 , f 1'..-
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, < J r 3 .;
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l ; i ! s .i i' l l 1 'i , a '
.
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- . . ._._, .,; a. .. ,_ ..;_,_ . : . _+
TEST CROSS REFERENCE f Page 4 SRO Exam PWR React or Organized by KA G oup PLANT WIDE GENERICS QUESTION VALUE KA 082 1.00 194001A102 004 1.00 194001A103 080 1.00 194001A103 079 1.00 194001A103 076 1.00 194001A106 081 1.00 194001A106 005 1.00 194001A111 003 1.00 194001K101 078 1.00 194001K102 002 1.00 194001K102 001 1.00 194001K102 008 1.00 194001K103 009 1.00 194001K103 007 1.00 194001K104 006 1.00 194001K104 075 1.00 194001K105 077 1.00 194001K114 PWG Total 17 00 PLNiT SYSTEMS Group I QUESTION VALUE KA 010 1.00 001000A106 011 1.00 001000K504 013 1.00 003000K201 012 1.00 003000K201 016 1.00 004000A207 015 1.00 004000K104 083 1.00 004000K404 014 1.00 004010A204 018 1.00 013000A301 033 1.00 013000A402 017 1.00 013000K103 084 1.00 014000A102 032 1.00 015000A303 031 1.00 015000K301 019 1.00 017020A201 020 1.00 022000A301 021 1.00 059000K419 024 1.00 061000A102 023 1.00 061000K103 022 1.00 061000K404
^
\ ( TESTCROSSREFERENCd Page 5 :
SRO Exam PWR Reaetor O'rganized by KA Group j l PLANT SYSTEMS I Group I QUESTION VALUE KA 1
: ......
PS-I Total 20.00 j Group II QUESTION VALUE- KA , 028 1.00 002000K106 029 1.00 002000K402 027 1.00 006000G004 026 1.00 010000A403 - 045 1.00 010000G013 065 1.00 010000K403 085 1.00- 011000A104 035 1.00 012000G008 034 1.00 012000K401 036 1.00 027000A401 _, 037 1.00 028000G005 030 1.00 033000K303 , 025 1.00 062000A401 + 039 1.00 064000A101 [ 040 1.00 064000A406 038 1.00 064000G007
'
......
'
PS-II Total 16.00 Group III c QUESTION VALUE KA ; 041 1,00 005000K201 044 1.00 007000K401, q 043 1.00 008000A301 046 1.00 000000K301-042 ' 1.00- 008030A30 ,
......
PS III Total . 00
......
...... .
'
PS Total 41.00
+
$
TEST CROSS-REFERENC Page 6-SRO Exam PWR Reactor Organized by KA Group EMERGENCY PLANT EVOLUTIONS Group I QUESTION VALUE KA 047 1.00 000001G003 086 1.00 000003G004 069 1.00 000003K309 094 1.00 000011A111 048 1.00 000011G010 087 1.00 000011K305 060 1.00 000011K313 062 1.00 000015G010 071 1.00 000015K102 063 1. 0 ]- 000024K301 064 1.00 000026K303 056 1.00 000029A209 090 1.00 000029G010 058 1 00 000040K101 055 1.00 000055G012 053 3.00 000055K302
*
073 .t 00 000057A217 074 1.00 000067G010 049 1.00 000069G012 089 1.00 000069G012 008 1.00 000074A207 052 1.00 000074K208-0S9 1.00 000074K305- -
.......
EPE-I Total 23.00 Group II QUESTION VALUE KA-066 1.00 000007A103 054 1.00 '000007A202 057 1.00 000008K302-050 1.00 00000U1012-
.068 1.00 000009K320- -093 1.00 000009K321 '059 1.00 000009K323 067 1.00 000025K101-096 1.00 -000038A139 095- 1.00 000038G012 061 1.00 000038K306 09 .00 000054G012 070 1.00 000054K304 100 1.00 000060K104l 091- -1.00 000061K101 y e
_ . _ . _ . . _ . _ _ _ _ . . _ . _ _ . _ _ _ _ _ _ . _ _ . . _ _ _ .
~. 1 -
TEST CROSS REFERENCU Page~ 7
!
i SRO Exam PWR Reactor I l 0rganized by KA Group i l EMERGENCY PLANT EVOLUTIONS Group II l QUESTION VALUE KA 097 1.00 000065K308 *
...... \
EPE-II Total 16.00 I Group III , QUESTION VALUE KA ,
!
098 1.00 000036A104 i 051 1,00 000056A203 .i 072 1.00 000056A247 ir
......
EPE-III Total 3.00 f
...... t
'
......
EPE Total 42.00 i
......
;
..... j
...... ;
Test Total 100.00 i i
!
i
!
,
l
. .
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