IR 05000302/1993301
| ML20058P887 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 12/03/1993 |
| From: | Hopper G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20058P882 | List: |
| References | |
| 50-302-93-301, NUDOCS 9312270319 | |
| Download: ML20058P887 (240) | |
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{{#Wiki_filter:cr - I' UNITED STATES .
. .##s nro% NUCLEAR REGULATORY COMMISSION yb ~ 'a REGION 11 " " '
i 101 MARIETTA STREET. N.W., SUITE 2900
- ti y ATLANTA, GEORGIA 303234199 . %, . . . . . / '
P.eport No.: 50-302/93-301 Licensee: Florida Power Corporation 3201 -34th Street, South St. Petersburg, FL 33733 Docket No.: 50-302 License No.: DPR-72 Facility Name: Crystal River 3 Examination Conducted: November 15-19, 1993 , Chief Examiner: I 3 eorge T. Hopper v Date Signed ; Accompanying Personnel: Richard Cain, INEL -; Approved By: b } Lawrence L. Lawyer, Chief e' Date Signed ! Operator Licensing Section Operations Branch Division of Reactor Safety ,
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SUMMARY
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Scope: NRC examiners conducted regular, announced operator licensing initial examinations during the period November 15-19, 1993. Examiners. administered ' examinations under the guidelines of the Examiner Standards (ES), NUREG-1021, Revision 7. Cne Senior Reactor Operator (SRO) candidate received both written and operating examinations. One SR0 candidate and one R0 candidate received ' written retake examinations onl '
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Results: Candidate Pass / Fail: SRO R0 Total Percent , Pass 2 1 3 100.0 % ! Fail 0 0 0 0.0 % ,
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No violations or deviations were identified, f
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9312270319 931203 PDR ADDCK 05000302 V PDR L _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
REPORT DETAILS- Persons Contacted Licensee Employees
* Boldt, Vice President, Nuclear Production' ;
*J. Carr, Nuclear Regulatory Specialist i L. Kelley, Director, Nuclear Operations Training :
*J. Lind, Manager, Licensed Operator Training t
*W. Marshall, Manager, Nuclear Plant Operations i
*J. Smith, Supervisor, License Training #
*J. Springer, Supervisor, Nuclear Simulator Training -i Other licensee employees contacted included instructors, engineers, technicians, operators, and office personne NRC Personnel ,
*R. Butcher, Senior Resident Inspector
*T. Cooper, Resident Inspector
*K. Landis, Chief, Reactor Projects Section 2B
*L. Lawyer, Chief, Operator Licensing Section ;
* Attended exit interview !
' Discussion
, Scope
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l NRC examiners conducted regular, announced operator licensing initial ) examinations during the period November 15-19, 1993._ Examiners ; administered examinations under the guidelines of the Examine l Standards (ES), NUREG-1021, Revision 7. One Senior Reactor Operator n (SRO) candidate received both written and operating _ examinations. One SR0 candidate and one R0 candidate received written retake examinations onl ,
b. _ Examination Development
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Facility representatives pre-reviewed the written, simulator and JPM l portions of the initial examination during the week of October 25, ' 1993. Their comments helped to improve and enhance the validity of-the examination. The success of this effort was demonstrated by the lack of any postexamination comments on the written examinatio ~
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Also, excellent support was provided_during the preparation visit to accurately and expeditiously validate _ the simulator scenarios and.JPM tasks. This was_ particularly appreciated since the prep week was done in conjunction with the administration of the requalificatio inspectio ]
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Report Details 2
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' Licensed Operator and Operator Candidate Performance
'I The candidate who received an operating examination exhibited a weakness in the area of emergency plan usage. The candidate experienced difficulty.in classifying events and.in the performance of protective action recommendation .
3. Written Test Review
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The following changes were made to the answer key following test administration: SR0 Question 18 - The answer was changed to "b". The distractors had been rearranged without changing the answer key. No substantive ; change to-the question was mad SRO Question 25 (R0 # 19) - Clarification made during the examination' ! by the proctor cianged the nature of the question resulting in item l
"a" being the correct answe SR0 Question 85 - Review of the question indicated that the correct answer for the given situation was "b", not "c".
4. Exit Interview .
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At the conclusion of the site visit, the examiners met with l representatives of the plant staff listed in paragraph one to discuss the
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results of the examinations. The licensee did not identify as , proprietary, any material provided to, or reviewed by the examiner ' Dissenting comments were not received from the license i l
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-j ENCLOSURE 2
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SIMULATOR FACILITY REPOR t f Facility Licensee: Crystal River Unit 3 , Facility Docket'No.: 50-302 !
Operating Tests Administered on: November 16, 1993
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This form is to be used only to report ot ervations. These observations do { not constitute audit or inspection findices and are not, without- further i 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 j future evaluations. No licensee action is required in response to these t observation j While conducting the simulator portion of the operating tests, the following - items were observed (if none, so state): 11EM DESCRIPTION l
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This document is removed from Official Use Only category on date of examinatio NRC Official Use Only w >
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U. S. NUCLEAR REGULATORY COMMISSION SITE-SPECIFIC WRITTEN EXAMINATION APPLICANT INFORMATION Name: Region: II Date: 1993/11/15 Facility / Unit: Crystal River 3 License Level: SRO Reactor Type: BW _ INSTRUCTIONS Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. The passing grade requires a final grade of at least 80 percent. Examination papers will be picked up 4 hours after the examination start All work done on this examination is my ow I have neither given nor received ai Applicant's Signature RESULTS Examination Value 100.00 Points Applicant's Score Points Applicant's Grade Percent
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'" ,SENIO;R REACTOR OPERATOR Page 2 ANSWER SHEET Multiple Choice (Circle or X your choice)
If you change your answer, write your selection in the blan MULTIPLE CHOICE 023 a b c d 001 a b c d 024 a b c d 002 a b c d 025 a b c d 003 a b c d 026 a b c d 004 a b c d 027 a b c d 005 a b c d 028 a b c d 006 a b c d 029 a b c d 007 a b c d 030 a b c d 008 a b c d 031 a b c d 009 a b c d 032 a b c d 010 a b c d 033 a b c d 011 a b c d 034 a b c d 012 a b c d 035 a b c 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 a b c d 043 a b c (1 021 a b c d 044 a b c d ___ 022 a b c d 045 a b c d i l l l 4 ..
h SENIOR REACTOR OPERATOR Page 3 ANSWER SHEET Multiple Choice (Circle or X your choice) If you change your answer, write your selection in the blank.
i 046 a b c d 068 a b c d l MULTIPLE CHOICE 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 l l
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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 j 067 a b c d 090 a b c d
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O'" , ,SENIO,R REACTOR OPERATOR Page 4 ANSWER SHEET Multiple Choice (Circle or X your choice) If you change your answer, write your selection in the blan a b c d MULTIPLE CHOICE 092 a b c d 093 a b c d 094 a b c d 095 a L 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 (********** END OF EXAMINATION **********) A
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4 % Page 5 NRC RULES AND GUID. TINES FOR LICENSE EXAMINATIONS During the administration of this examination, the following rules apply: Cheating on the examination will result in a denial of your application and could result in more severe penaltie . After you complete the examination, sign the statement on the cover sheet indicating that the work is your own and you have not received or given assistance in completing the examinatio . To pass the examination, you must achieve a grade of 80 percent or greate . The point value for each question is indicated in parentheses after the question numbe . There is a time limit of 4 hours for completing the examinatio . Use only black ink or dark pencil to ensure legible copie . Print your name in the blank provided on the examination cover sheet and the answer shee . Mark your answers on the answer sheet provided and do not leave any question blan . If the intent of a question is unclear, ask questions of the examiner onl . Restroom trips are permitted, but only one applicant at a time will be allowed to leave. Avoid all contact with anyone outside the examination room to eliminate even the appearance or possibility of cheatin . When you complete the examination, assemble a package including the examination questions, examination aids, and answer sheets and give it to the examiner or proctor. Remember to sign the statement on the examination cover shee . After you have turned in your examination, leave the examination area as defined by the examine ; _
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,m , - , SENIOR REACTOR OPERATOR Page 7 QUESTION: 001 (1.00)
Given the following conditions:
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Make-Up and Purification Pump, MUP-1A, quarterly surveillance is in progres RCP Seal Injection Flow Indicator, MU-27-FI, indicates 35 gpm and flow needs to be adjuste RCP Seal Injection Flow Control Valve, MUV-16, is Blue-tagge The Blue Tag instructs the operator NOT to adjust MUV-16 due to packing leakag Select the operator action required to complete the surveillanc Stop the surveillance and contact the Nuclear Shift Superviso Read the Blue Tag and continue with the surveillanc Send an operator to monitor MUV-16 and continue with the surveillanc Clear the Blue Tag, continue with the surveillance, and notify the Nuclear Shift Supervisor of action taken when the surveillance is completed.
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.QUESTION:
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Given the following conditions:
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A 21 year old male radiation worke A current NRC Form 5 on file for the worker.- 'i
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He has received 47.0 rem to the skin of'the whole body this year (SDE-WB)
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dose is ;
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According to 10CFR20.101, what is the MAXIMUM amount of Total Effective Dose Equivalent (TEDE) he can receive for the remainder of the year? j
, .5 re i .0 re .0 ren, .0 re l t
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QUESTION: 003 (1.00) I Which plant communication system is dedicated for fire reporting? a. PL-2 on the. PAX Syste . b. Channel 5 on the hand held radio I c. PL-1 on the PAX Syste f d. Channels 13 and 14 on the hand held radio j l l I
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A reactor heat-up/startup is in progres The plant is in Mode Choose the MINIMUM number of Licensed Reactor Operators (ROs) and their required location for the above condition RO located inside the control roo RO located at the main control boar c. 2 Ros, both inside the control room and one located at the main control boar d. 2 ROs, both inside the control room and both located at the main control boar QUESTION: 005 (1.00) The Nuclear Shift Supervisor (NSS) directs you, with face to face communication, to line up the Make-Up and Purification Pump, MUP-1B, for the annual surveillanc What communication requirement is NOT necessary for this exchange? The use of: a. repeat backs or paraphrasing the messag b. title /name identifier c. confirmation from tne NS d. proper names and equipment number . _ _ _ . , . . _ .._ __ _ . . ._
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A small' break loss of coolant accident (LOCA) is in 1 progres . i
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The Nuclear Shift Supervisor has declared'an Alert and j announced it over the plant paging syste *
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You are the On-Shift Nuclear Auxiliary Operato Where should you ! report to upon hearing the declaration of Alert? .)
: The Technical Support Center (TSC)
{ The Operational Support Center (OSC) !
. The Operations Conference Room !
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QUESTION: 007 (1.00) j
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A General Emergency has been declare j
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A radioactive release is occurring from the Auxiliary .l * Buildin A Padiochemist has injured her leg and cannot exit the ; Auv.liary Buildin !
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Choose.the requirements for an emergency worker to. assist the ; radiochemis .l The worker must'be a male and have his dose limited to-75 rem ; whole bod ! I The worker can be male or female and have his or her dose i limited to 25 rem whole bod j The worker must be a volunteer and have their dose-limited to -- 75 rem whole bod The worker can be male or female and his or her exposure should i not exceed quarterly exposure limit j i i
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The plant has been in Mode The Letdown and Purification System has had extensive work performed on i A valve line-up on the Letdown and Purification System inside the Reactor Building is require Select your responsibilities as a radiation worker while performing the valve line-up, Decline to do the job because the dose rates are to hig Perform the job quickly and safely to reduce the time spent in r the radiation are Perform the job by verifying the valve positions using the previously completed valve line-u Perform the job from a distance by using a flashlight and direct line of sight to observe valve positio QUESTION: 009 (1.00) You are performing an independent verification (2nd checker) of a tagout on the High Pressure Injection System. The "First Person" clearing the tagout asks you to observe him repositioning a throttle valv Select the Independent verifier (2nd checker) respons This is appropriate because it is a throttle valv This is incorrect because it is NOT " Independent Verification." This is appropriate because it is restoration of a non-safety system componen This is incorrect because Independent Verification is required to be a " hands on" performanc E-
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Breathing air is Jined up to the Reactor Building for use by the mechanics to replace a Reactor Coolant Pump (RCP) Sea What prevents the isolation of the breathing air supply valve? The Breathing Air Supply valve is: red tagged ope blue tagged ope white tagged ope locked open with no ta QUESTION: 011 (1.00) Given the following conditions:
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Reactor Building (RB) walkdowns are being performed prior to RB close ou An operator performing a walkdown has been informed that his annual dose is 3.950 rem and that his yearly ADMINISTRATIVE limit must be increase What is the operator's NEW limit and who must approve the increase? The Radiation Protection Manager can increase the limit to 4.000 re The Director, Nuclear Plant Operations can increase the limit to 4.000 re The Radiation Protection Manager can increase the limit to 5.000 re , The Director, Nuclear Plant Operations can increase the limit to 5.000 re .
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Refueling operations are in progres New fuel is in the transfer carriage and traveling ; from the spent fuel pool to the reactor buildin j
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The SRO and RO monitoring refueling operations are on:the i Main Fuel Handling Bridg ;
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A Nuclear Operator (NO)'comes up to the bridge with a tag out requiring 1 an independent verification on a tag located a short distance awa Select the allowed action of the refueling operators (RO & SRO). l
' The Refueling RO verifies the tag since the Refueling SRO can NOT have any other concurrent dutie ;
6 The Refueling SRO verifies the tag since the Refueling RO is -l NOT allowed to leave the Main Fuel Handling Bridg ji Neither the Refueling SRO or RO verify the tag and inform-the' , No that he must'use another operator for verificatio The Refueling RO verifies the tag after being relieved by the l Refueling SRO on the Main Fuel Handling Bridg j i r I
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A reactor trip has occurre The trip was due to a fault on one of the Main Generator output breaker The Shift Supervisor has directed you to make up the tagout for the work on the Main Generator output breakers. What must be completed prior to hanging the in-plant tagout? A system clearance from the Dispatcher to the on duty Shift Supervisor must be in plac A grounding device must be installed on the main generator side of the main generator output breaker The Emergency Diesel Generators must be tested and one left running and loaded on a 4160 bu The Main Generator must be purged of hydrogen and filled with nitrogen prior work on the output breaker ; i
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-QUESTION: 014 (1.00) l l i Given the following conditions: --
The quarterly Reactor Building Spray Pump surveillance is [ in progres ; p -- The operator who is performing the surveillance and an I&C i l technician inform the Shift Supervisor that an out of j l tolerance reading listed in the surveillance is incorrec The I&C technician states that a new flow instrument has i been installed and the tolerances should be plus or minu ! 5% and not 2% as indicated on the procedur j
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Select the action to be take ! Stop the surveillance and perform a Temporary Procedure Change l per AI-400 l l Continue with the surveillance and notify plant managemen j Continue with the surveillance and record the deviation in the j plant lo i Stop the surveillance and perform a Permanent Procedure , Revision per AI-400 l
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QUESTION: 015 (1.00) Given the following conditions:
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A start-up is in progress following a refueling outag A Chief Nuclear Operator reports that the posted Axial Power Imbalance Graphs are for the previous fuel cycl Choose the required shift supervision actio Re- tve the previous fuel cycle information and replace it with tia current fuel cycle information, Place the current fuel cycle information adjacent to the previous fuel cycle informatio Stop the start-up until the Director of Nuclear Plant Operations (DNPO) approves the new fuel cycle inforaatio Place one line through all the previous fuel cycle information and label it Not Applicable (N\A).
QUESTION: 016 (1.00) Emergency Feedwater Initiation and Control has initiated and is creating an overcooling condition in the Reactor Coolant System (RUS). What is required before Emergency Feedwater can be throttled? Chief Nuclear Operator's permission and a B&W guideline documen The Shift Supervisor's permission and a temporary test procedur The Assistant Shift Supervisor's permission and an approved plant procedur The Director, Nuclear Plant Operations permission and a NRC guideline documen . O
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QUESTION: 017 (1.00) Given the following conditions:
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The Plant is in Mode Decay Heat Pump 1A, DHP-1A, is runnin The LOAD Dispatcher has informed the Control Room that some breaker work will be performed in BOTH switchyards on this shif What is required to be completed prior to performing the breaker work? The Job Supervisor's permission and swapping the supply power of DHP-1A to the other switchyard suppl Manager-on-Call \ Nuclear Shift Manager's permission and Decay Heat "A" train components powered from the Emergency Diesel Generator, Manager-on-Call \ Nuclear Shift Manager's permission and Decay Heat "B" train components powered from the Emergency Diesel Generato The Shift Supervisor's permission and swapping the supply power of DHP-1B to the other switchyard suppl QUESTION. 018 (1.00) Given the following conditions:
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The Reactor Bailey and Diamond station are in manua All other ICS stations are in automati A " Reactor Limited by Feedwater" crosslimit alarm is illuminate Select the indication on the neutron error mete The meter indicates plus 2%. The meter indicates 0%. The meter indicates minus 1%. The meter indicates minus 2%.
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QUESTION: 019 (1.00) Given the following conditions:
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A Reactor Startup is in progress, ECP is 30% on Group Group 7 rods are being withdrawn and are at 28%.
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Rod 7-5 drops and its "in-limit" lamp is illuminate Start-up rate indicates NEGATIVE 0.10 dpm, counts are decreasin The Reactor Operator has stopped withdrawing control rod Choose the action required to recover the dropped ro Determine the cause for the dropped rod: and recover the dropped rod by withdrawing it to the grou and drive Group 7 rods to the in-limit, relatch all Group 7 rods, and continue with startu and continue the reactor startup until Group 7 rods are at 30%, then recover the dropped ro and at a minimum, fully insert rod groups 5, 6, and 7 and reverify initial conditions of OP-21 , _ _ _ _ . . . _ . _ _ _
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Reactor power is 100%.
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An approved, special Reactor Protection System (RPS) test is in progres During the test, " Electronic trips E & F" are actuated simultaneousl ;
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Channel trips A, B, C, and D are normal (NOT TRIPPED). I Select the expected Control Rod Drive (CRD) respons l All rods. fully insert (Drop). Regulating Rod Groups 5, 6, and 7 insert (Drop). Safety Rod Groups 1, 2, 3, and 4 insert (Drop).
, Regulating Rod Groups 5, 6, and 7 remain in position but cannot be move ; t I:
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QUESTION: 021 (1.00) .! Given-the following conditions:- q i
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The reactor is at 250 EFP It has been 14 hours since a reactor trip has occurre Prior to the trip, the reactor had been at 100% power for ' 30 day You are instructed to perform a reactor start-u A 2 hour delay has
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What does the operator need to watch for and why? The operator needs to monitor for the: a. ,MINUS 1.0% delta k/k error margin due to xenon deca PLUS 1.0% delta k/k error margin due to xenon' buildu PLUS 1.0% delta k/k error margin due to samarium buildu MINUS 1.0% delta k/k error margin due to samarium deca QUESTION: 022 (1.00) Given the following conditions:
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Reactor Coolant Pump 3B2 (RCP-1D) Controlled Bleed Off flow l has decrease RCP-1D Seal leakoff flow has increase RCP Seal Injection Flow is constan Select the cause for the abnormal indication First stage RCP seal degradation is occurrin Third stage RCP seal degradation is occurrin l RCS letdown has been increase I RCS pressure has decrease ') l
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QUESTION: 023 (1.00) Given the following conditions:
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Reactor power is 100% and stabl All ICS stations are in " automatic."
-- Control rods have been moving INWARD over the past 24 hour The RCS leak rate has DECREASED over the past 24 hour Tave and pressurizer level have been constan Select t.he cause for these changes, Xenon has been increasing to reach equili;.,riu Inadvertent boration is occurrin Inadvertent boron dilution is occurrin Samarium has been decreasing to reach equilibriu QUESTION: 024 (1.00) Given the following conditions:
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Reactor power is 90% and stabl The Reactor Demand station is in " manual."
-- ICS is in " track."
-- The remaining ICS stations are in " automatic."
-- A 3 ppm boron dilution is performe minutes has elapsed since the dilution with NO additional operator actions take Which one of the following describes Tave's expected response? Tave will: vary and stabilize at a higher valu decrease and stabilize at a lower valu vary and return to the same value, decrease and then increase to a higher valu :
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Reactor Coolant System (RCS) pressure 1490 psi Reactor Building pressure 3.5 psi MUP-1A, Make-Up pump 1A Runnin MUP-1B Selected to B ES-416 MUP-1C In ES standb Select the mode of operation for the Make-Up Pumps, MUP-1A is running MUP-1B is running hN i MUP-1C is running A8 k MUP-1A is running ~ MUP-1B is running r ixM r MUP-1C is off A g, C U_] ~@ MUP-1A is off - -- MUP-1B is running Ske MUP-1C is running Here- MUP-1A is running MUP-1B is, off MUP-1C is running k [ i ofv // d
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-QUESTION: 026 (1.00)
Given the following conditions:
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RCS pressure 1625 psig
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Reactor Building pressure 0.50 psig
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A plant heatup is in progres i-What must be done to the ESAS PRIOR to exceeding 1650 psig in the RCS? l Place LPI in ES Standb ' Place HPI in ES Standb Open both Core Flood Outlet Valves and lock open the breaker ,
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QUESTION: 027 (1.00) ; i i Given the following conditions:
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NI-5 power 102.5%
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NI-6 power 97.5%
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Heat balance power 100%. , i Select the NI that REQUIRES calibration and the reason wh ! NI-5 needs calibration in order to correct average NI power, which in turn, would allow greater megawatt power outpu NI-5 needs calibration because a 2% nuclear instrument error was assumed in the Safety Analysis Repor NI-6 needs calibration in order to-correct average NI power, which in turn, would allow greater megawatt power outpu ;
-t' NI-6 needs calibration because a 2% heat balance. error was assumed in the Safety Analysis Repor l t
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Page 24 . . k QUESTION: 028 (1.00) Given the following conditions:
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A station blackout has occurre Natural Circulation has been establishe The plant is in " Hot Standby."
-- Subcooling Margin is 45 degrees The Shift Supervisor directs the Chief Nuclear Operator #1 (CNO 1) to verify the subcooling margin with the use of the Steam Table Select the instrumentation tne CNO 1 will use for this verification, RCS T-cold leg tempe:wiure and RCS narrow range pressur RCS T-hot leg tempere,ture and RCS wide range pressur Core exit thermocouples and RCS wide range pressur Core exit thermocouples and OTSG saturation pressur QUESTION: 029 (1.00) The Reactor Building Cooling and Isolation System will limit the post accident ambient pressures to design value What is the MINIMUM combination of Reactor Luilding (RB) Spray System Trains and Emergency Cooling Units AHF-1A, -1B, and -1C needed to meet this design criteria? ONE RB Spray Train and ONE Emergency Cooling Uni NO RB Spray Trains and TWO Emergency Cooling Unit ONE RB Spray Train and NO Emergency Cooling Unit NO RB Spray Trains and THREE Emergency Cooling Unit c e'"' Page 25
.SEMIQR REACTOR OPERATOR . .
QUESTION: 030 (1.00) Given the following conditions:
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A Loss of Offsite Power has occurre ESAS has actuated on 1500 psig RCS pressure decreasin What is the power supply to the motor driven Emergency Feedwater Pump? Emergency Diesel Generator 1B Unit Bus 3 Emergency Diesel Generator 1 Unit Bus 3 QUESTION: 031 (1.00) Given the following conditions:
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A reactor trip has occurred from 100% powe Concurrent with the reactor trip, a loss of both Main and Emergency Feedwater has occurre Select the required EOP section to be entered for these condition EOP-02, " Vital System Status Verification." EOP-04, " Inadequate Heat Transfer." EOP-05, " Excessive Heat Transfer." EOP-07, " Inadequate Core Cooling."
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. . QUESTION: 032 (1.00) Given the following conditions:
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Refueling operations are in progress inside the Reactor Buildin A continuous Reactor Building purge is in progres RM-Al is required to be removed from service for calibratio Select the required operator action when RM-Al is removed from Servic Isolate containment purg Take a grab air sample every 4 hours, and continue containment purg Bypass the RM-Al interlocks and place a temporary radiation monitor on the refueling dec Require all personnel inside the reactor building to wear respirator QUESTION: 033 (1.00) Given the following conditions:
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A reactor trip has occurred from 100% powe " Radiation High Alarm,"is illuminated for Radiation Monitor RM-L RM-G6 and RM-G10, Make-up Tank and Pump Room radiation monitors indicate increasing radiation level The Make-up Pre-filter High " Delta p" Alarm is illuminate Select the cause for these indication Failed fuel has occurred in the reactor core, The make-up demineralizers have faile There is a RCS leak in the make-up tank roo The reactor trip has caused a crud burs . _ _ _ _ _ _ _ , _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . . _ _ _ _ _ _ _ _ . _ _ .
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. SENIOR REACTOR OPERATOR Page 27
. . QUESTION: 034 (1.00) Reactor Building pressure is 3.0 psig and increasin The Shift Supervisor, anticipating 4.0 psig building pressure, has you manually acteate the LPI system by depressing both pushbutton What mode of operation are the LPI pumps in? The LPI pumps are: running and on recirculation until RCS pressure lowers to 500 psi running and flow is being injected into the RC idle (NOT running), because the "HPI seal in" is NOT present, idle (NOT running), until the LPI system valves realign, and then they receive an auto-start signa QUESTION: 035 (1.00) Given the following conditions:
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RCS pressure is 2205 psig and increasin The Pressurizer Spray Control valve " green" indicating light has extinguished and one amber light has illuminate A second amber light on the Pressurizer Spray Control valve has illuminate Choose the cause for the illumination of the second amber ligh The Pressurizer Spray valve is 100% ope The Pressurizer Spray Block valve is closed, The Pressurizer heater banks "A" and "B" are off, The Pressurizer Spray valve is 40% ope _ - _ ___ _ _ _ . .
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QUESTION: 036 (1.00)
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Given the following. conditions:
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The Pilot-Operated Relief Valve (PORV/RCV-10) is in
" automatic." ! --
RCS pressure has increased to the POR" automatic open- ! setpoin The PORV " red" and " amber" lights are' illuminate l
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-Select the status of the PORV/RCV-10 in this mod !
.The PORV:
+ is 40% ope !
'i solenoid valve is energize ;
i is 100% ope i is " interlocked" close <
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> QUESTION: 037 (1.00)
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i-Given the following conditions: !
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A reactor trip'from 75% power has occurre !
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A malfunction in Rapid Feedwater Reduction (RFR) circuitry : has caused main feedwater to overfee l*
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Pressurizer heater banks "D" and "E" have been placed in "on" but they did NOT energiz j
. Select the cause for the lack of power to Pressurizer heater banks "D" ., and "E". ; An insurge into the pressurizer has caused RCS pressure'to-increase to 2155 psi The Pressurizer Spray valve "Open Interlock" has removed power from pressurizer heater bank , The Pilot-Operated Relief Block Valve (PORV) is closed and is 1 interlocked with the pressurizer heater ; An outsurge from the pressurizer has caused level to decrease !
below 40 inche l
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,SENIQR REACTOR OPERATOR Page 30 . .
QUESTION: 038 (1.00) Given the following conditions:
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Pressurizer level and pressure control are in " automatic."
-- A small leak (0.20 gpm) develops in the pressurizer steam spac Select the expected Pressurizer and Make-up Tank level respons Pressurizer level remains constan Make-up Tank level decrease Pressurizer level increase Make-up Tank level decrease Pressurizer level increase Make-up Tank level increase Pressurizer level decrease Make-up Tank level increases QUESTION: 039 (1.00) Which condition is used to indicate to RPS that the Main Turbine is tripped? a. The EHC oil pressure is less than or equal to 45 psig, b. The Auto Stop oil pressure is less than or equal to 45 psi The Turbine Throttle valve LVDTs indicate close The Turbine Governor valve LVDTs indicate closed.
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. . QUESTION: 040 (1.00) Given the following conditions:
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NI-5 has failed and Channel "A" Reactor Protection System (RPS) is in the " Trip Condition."
-- The I&C technician must perform NI-6, Channel "B" RPS, testin A " Minimum Degree of Redundancy" of ONE must be maintaine What must be done with RPS Channels "A" and "B" in order to perform Channel "B" testing? Place Channel "B" RPS: in " bypass" and Channel "A" RPS in " bypass." in " bypass" and leave Channel "A" RPS " tripped." in " shutdown bypass" and leave Channel "A" RPS " tripped." in " Normal" and leave Channel "A" RPS " tripped."
QUESTION: 041 (1.00) Given the following conditions:
--
All ICS stations are in " automatic."
-- Reactor Power is 100%. It is deshed to adjust power imbalance using the Axial Power Shaping Rods (APSRs). What is required to operate the APSRs? a. The Diamond Controi Station must be placed in " Manual."
b. " Sequence override" must be selected at the Diamond Control Statio c. The " Single Select Switch" must be in "All."
d. The " Group Select Switch" must be in the " Group 8" positio L - l j ,SENIQR REACTOR OPERATOR O O Page 32
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. QUESTION: 042 (1.00)
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A reactor trip has occurre .
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The cause has been determined and correcte l
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Rod Groups 1 thru 7 are being latched prior to withdrawa Groups 1 through 4 have been latche l l
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Group 5 is selected and CANNOT be latche . l l Select the cause for the Inability to latch Group 5 rod The " Manual Dilute Permit" interlock requirements have NOT been me ; The Group position indication has NOT been reset to zer Safety Rod Groups 1 through 4 are NOT at their out-limi The " Auxiliary" power supply is De-energize . ' l
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. QUESTION: 043 (1.00) 1
1 Given the following condition ,
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An electrical fault has caused erratic indication in -l Channel "A" RP l
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The SASS monitors are NOT operationa I
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RCS pressure is mimicking the Channel "A" RPS indication Select the cause'for RCS pressure fluctuations, The Non-Nuclear Instrumentation (NNI) control signal is- I selected to Channel "A" RP ) i The Nuclear Instrumentation (NI) Channel 5 signal is supplying IC c. The RCS flow signal supplied to the Main Control Board (MCB) is selected to Channel "A" RP d. The RCS flow signal supplied to the ICS is selected to Channel
"A" RP !
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,SENIQR REACTOR OPERATOR Page 33
. . QUESTION: 044 (1.00) The selected Main Steam Header Pressure instrument is degrading (failing) from 895 psig to zero over 5 minute Choose the SASS response to this failur The SASS channel: trips and swaps control to the non-changing signa trips and the amber mismatch lamp is illuminate reverts to the " manual" mode and the amber mismatch lamp is illuminate reverts to the " manual" mode and the red trip lamp is illuminate QUESTION: 045 (1.00) Given the following conditions:
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Spent fuel temperature is increasin Boron concentration is constan Nuclear Services Surge Tank, SWT-1, has a low-low level alar Select the cause for the above indication The Spent Fuel Pool; has an inadequate shutdown margin (SDM). ventilation has been secure level is decreasin cooling has been los ,SENIQR REACTOR OPERATOR e
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Page 34 i i . . 1 QUESTION: 046 (1.00)
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Given the following conditions:
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A loss of decay heat removal has occurre The RCS is NOT exposed to atmosphere (Filled and vented).
-- RCS incore temperature 200 degrees RCS pressure 240 psi Select the method to remove decay hea An OTSG with a Reactor Coolant Pump Feedwater (Emergency or Main) Bypass valves (Turbine or Atmospheric). Spent Fuel Cooling System with a valve line-up to supply the Decay Heat Syste The BWST using a gravity feed metho The BWST using the Reactor Building Spray pump QUESTION: 047 (1.00) The 480 VAC ES MCC supplying an inverter has been de-energize Select the voltage input to the " Static Switch" supplied by the inverte VDC VAC VAC VDC
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. . QUESTION: 048 (1.00) Given the following conditions:
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The "A" Emergency Diesel Generator (EDG) is running and supplying loads on the 4160 ES Bu Panel DPDP-6A has been de-energized, which causes a loss of DC control power to "A" ED Select the resultant operating condition of the "A" ED The "A" EDG: , trips with no alarm indication trips and all attendant alarms operate as require remains running with the shut down relay defeated, remains running with the emergency stop relay defeate , F
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,SENIQR' REACTOR OPERATOR Page 36
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-QUESTION: 049 (1.00) '
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Given the following conditions: t
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Refueling operations are in progres '
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A new fuel assembly is being transferred from the spent ! fuel pool to the fuel transfer cana r
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The running "A" Decay Heat pump must be secured for i maintenance prior to_ placing the "B" Decay Heat train in ? serv 1C r Choose the evolution which should NOT be performed while the decay heat ! train is of :l t The transferring of New fuel from the Spent Fuel Pool to the ; Fuel Transfer Cana ' Increasing the level of the Spent' Fuel Pool by adding de- ,
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ionized wate Increasing Decay-Heat flow above the minimum continuous DH pump ; flow of 1400 gp ;
' Lowering the Reactor Vessel level during refueling operation !
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. . ; i . QUESTION: 050 (1.00)
. -Given the following conditions:
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The RCS has been drained down to "MID LOOP OPERATIONS." l
-- "A" Decay Heat Train is in operation at 1400 gp Maintenance is to be performed on Bus "ES A 4160." .
i Select the required mode of operation for the Decay Heat Syste , Isolate both Decay Heat trains until EGDG-1A output breaker , maintenance is complet ! Cross-tie the 4160 VAC ES Bus 3A with 4160 VAC ES Bus 3 Place the "B" Decay Heat Train in service and the "A" Decay Heat Train in standb i Switch the breaker control handle for DHV-3, Inboard Decay Hea !
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Suction Isolation, to the closed positio !
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QUESTION: 051 (1.00) 3
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Given the following conditions: j
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Reactor Building pressure is 4 psi The Reactor Coolant Drain Tank (RCDT) rupture disc has been rupture A Pressurizer Code Safety valve is liftin Pressurizer Temperature is'656 degrees What is the tail pipe temperature for the lifting safety? f a. 158 degrees .) l b. 212 degrees j i c. 230 degrees' .j
d. 656 degrees I
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. . QUESTION: 052 (1.00) Given the following conditions:
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Instrument air pressure has decreased to 83 psig and is constan VP-580, " Plant Safety Verification Procedure" is being performe A " differential flow" alarm on the Control Rod Drive Mechanisms REQUIRED isolation of the Nuclear Services Closed Cycle Cooling System (NSCCCS).
Select the REQUIRED operator action once the NSCCCS is isolate Trip the reactor and concurrently perform EOP-02, " Vital System Status Verification." Stop all 4 Reactor Coolant Pump Trip both Main Feedwater Pump Isolate Instrument Air Valve IAV-30 (cross-connect between Instrument Air and Service Air).
QUESTION: 053 (1.00) Given the following conditions:
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Reactor power is 100%.
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Group 7 rods, the controlling group, begins to move IN continuousl Rod motion stops when group 7 is at their in-limi Proper rod sequencing was observe Reactor power is 65%. What core operating limit is closest to exceeding its limit? a. Quadrant Power Tilt b. Regulating Rod Insertion Limits c. Rod Program d. Axial Power Shaping Rod Insertion Limits
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. . QUESTION: 054 (1.00) Given the following conditions:
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Group 7, Rod 5 sticks while withdrawing rods for criticalit An asymmetric rod fault lamp is illuminated on the Diamond Pane Select the position of Group 7, Rod 5 when the fault lamp illuminate a. Exactly 9 inches misaligned from another rod in Group b. Exactly 6.0% index difference from another rod in Group c. Greater than 9 inches misaligned from the Group 7 averag d. Greater than 9% index difference from another rod in Group QUESTION: 055 (1.00) Given the following conditions:
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An Asymmetric Rod Fault alarm is illuminate The out-Inhibit Lamp is illuminate Group 7 rod 3 rod bottom light is illuminate Choose the " operating condition" of the IC a. A plant runback is in progress to less than 60% powe b. The Diamond Panel has shifted to " manual" and the Auto-Inhibit light is illuminate c. The Diamond Panel has shifted to " manual" and the Sequence-Inhibit light is illuminate d. Feedwater-Reactor crosslimits are in effect causing a plant runback to less than 60% powe __ _ _ - _ _ - _ _ _ - _ _ _ _ _ _ _ _ _ _ - _ _ _
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. . QUESTION: 056 (1.00) F Given the following conditions:
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Group 7, rod 3 (Rod 7-3) had been previously stuc Rod 7-3 has been freed using procedure OP-50 '
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However, the relative rod position indication (RPI) on the Position Indication (PI) panel for Rod 7-3 DOES NOT agree with its absolute position indication (API).
What is required to realign Rod 7-3 RPI to match Rod 7-3 API? Move Group 7 rods to Rod 7-3 API and realign all Group 7 rods with Rod 7-3 AP Drive all Group 7 rods to the in-limit and realign the rods to the Zero positio Withdraw all Group 7 rods to the nearest Zone indicating lamp and realign all Group 7 rods to the zone reference indicatio Select Rod 7-3 on the Group and Single Select switch, and use the Roset Pulser to align RPI with AP QUESTION: 057 (1.00) Given the following conditions:
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A small break LOCA has occurred which CANNOT be isolate RCS pressure 650 psi Subcooling Margin (SCM) 52 degrees Select the required status for the Core Flood Tank The Core Flood Tanks should be: isolate b. vente c. allowed to inject until SCM is greater than 70 degrees allowed to inject until LPI injection into the core is verifie ,n
' . SENIOR REACTOR OPERATOR Page 41 . .
QUESTION: 058 (1.00) Select the plant parameters which indicate a LOCA is in progress PRIOR to a reactor trip being initiated.
, RCS Pressure is decreasin RCS Temperature is decreasin Secondary Saturation Temperature is decreasin RCS Pressure is decreasin RCS Temperature is constan Secondary Saturation Temperature is constant, RCS Pressure is decreasin RCS Temperature is increasin Secondary Saturation Temperature is decreasin RCS Pressure is constan RCS Temperature is decreasin Secondary Saturation Temperature is constan QUESTION: 059 (1.00) Given the following conditions:
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The reactor has tripped from 100% powe The Subcooling Margin is O degrees RCPs are of Primary T-hot and Primary T-sat are equa Primary T-cold and Secondary T-sat are equa Select the mode of RCS cooling occurring for the present condition Single phase Natural Circulation Forced convection Boiler-condenser Natural Circulation Natural Conduction i
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. . QUESTION: 060 (1.00) Given the following conditions:
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A slow seal degradation is indicated on "A" RC "A" Reactor Coolant Pump (RCP) Third Stage Seal temperature is 175 degrees Choose the required IMMEDIATE operator actio a. Stop the "A" RC b. Reduce reactor power to less than 72% and stop the "A" RC c. Verify the high temperature condition and that RCP cooling water and seal injection is adequat d. Trip the Reactor, isolate the Reactor Building, and immediately stop the "A" RC QUESTION: 061 (1.00) Given the following conditions:
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Due to a malfunction on one RCP, the plant is in "Three Pump" operation The proper precautions have been taken and the plant has returned to 75% powe Select the OTSG level for the RCS loop which has 2 RCPs operatin OTSG 1evel is equivalent to the 4 pump: a. 25% power leve b. 50% power leve c. 75% power leve d. 100% power leve l I l
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QUESTION: 062 (1.00) Given.the following conditions:
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A loss of Subcooling Margin has occurre Reactor Coolant System (RCS) pressure is 400 psi EXCEPT for Reactor Building Spray / Chemical Addition, all f Emergency Safeguards Actuation Systems (ESAS) hav initiate , Choose the ESAS component (s) allowed to be repositioned out.of their .i ESAS required positio The RCP seal return and SW cooling valves to the running RCP I One HPI pump can be stoppe EFIC can be bypassed and OTSG levels maintained at 50% in the j operating rang If LPI flow is less than 500 gpm, then stop both LPI pumps, l QUESTION: 063 (1.00) f i The Diverse Scram System (DSS) provides a reactor trip signal as a back' [ up for the Reactor Protection System (RPS). i
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Select the INSTRUMENT and PARAMETER used for the input to the DS :
! RPS Channels A & B reactor coolant temperature i
! Remote Shutdown Panel reactor coolant pressure ; Emergency Safeguards Actuation System reactor building pressure RPS Channels C & D High Power / Flux ;
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S'" Page 44 . . QUESTION: 064 (1.00) Given the following conditions:
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Reactor power is 28%.
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Main Feedwater has suddenly decreased to 15% feed flo Select the expected automatic plant response, The Reactor is Feedwater Limited (crosslimits) and feedwater flow is increased to 28%. The Feedwater is Reactor Limited (crosslimits) and feedwater flow is increased to 28%. The operating Main Feedwater pump shifts from speed control to
" delta p" contro EFIC initiates and the Main Turbine is trippe QUESTION: 065 (1.00)
The function of the ATWS Mitigation System Actuation Circuitry (AMSAC) is to provide: a diverse means to initiate EFIC and trip the main turbine upon a loss of feedwater during power operatio b. a reactor trip signal independent of the Reactor Protective System on indication of high RCS pressur c. a diverse means to balance reactor power and Main Feedwater flow during power operatio d. a reactor runback signal independent of the Integrated Control System on indication of RCS flow mismatc P e'"
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QUESTION: 066 (1.00) Given the following conditions:
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A large steam leak has occurre OTSG "A" indicates 590 psi OTSG "B" indicates 460 psi Select the expected EFIC feedwater respons EFIC vector logic will send: open commands to both OTSG feedwater valves and will feed both OTSG a close command to "A" OTSG feedwater valves, an open command to "B" OTSG feedwater valves, and will feed only "B" OTS an open command to "A" OTSG feedwater valves, a close command to "B" OTSG feedwater valves, and will feed only "A" OTS close commands to both OTSG feedwater valves and will NOT feed either OTS ,\. .p Q
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. ..- r QUESTION: 067 (1.00)
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Given.the'following conditions:
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An " Excessive Heat Transfer" transient has occurre il What criteria is used to determine a " Stable OTSG7"
! OTSG pressure is decreasing ",
OTSG pressure is at Psat for the RCS Tc OTSG level is constant >; e OTSG pressure is NOT decreasing j OTSG pressure is at Psat for the RCS Tc - OTSG level is constant i
; OTSG pressure is NOT decreasing OTSG pressure is 100 psi below Psat for-the RCS Tc ,
OTSG 1evel is increasing OTSG pressure is decreasing from the OTSG safety. lift setpoint-OTSG pressure is at Psat for the RCS-Tc ; OTSG level is increasing {
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QUESTION: 068 (1. 0 0) i Given the following conditions: l
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A station blackout has occurre l Select the equipment responsible for cooling the reactor cor 'a-3 The Make-up pumps used as HPI pump The Electric Driven Emergency Feedwater pum J
'1 c. The Decay Heat pumps used as LPI pump ;!
) The Steam Driven Emergency Feedwater pum j
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SENIOR REACTOR OPERATOR Page 47 . , QUESTION: 069 (1.00) Given the following conditions:
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Reactor refueling is in progres EGDG-1A is out of service for maintenanc A Loss of Off-site Power has occurre EGDG-1B FAILED to star Choose the AP/EOP to be entere EOP-12, " Station Blackout." EOP-11, " Loss of Decay Heat Removal." AP-581, " Loss of NNI-X." EOP-4, " Inadequate Heat Transfer."
QUESTION: 070 (1.00) Given the following conditions:
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Reactor power was at 100%.
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A station blackout has occurre A loss of subcooling margin has occurre Select the applicable MAXIMUM RCS cooldown rat a. The maximum achievable rate Less than or equal to 100 degrees F per hour c. Less than or equal to 50 degrees F per hour d. Less than or equal to 25 degrees F per hour
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i l l QUESTION: 071 (1.00) l l l Given the following condition i
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The NNI-X white indicating light on the Redundant , I Instrument Panel has extinguishe Annunciator alarm, "NNI-X Power Failure," is illuminate A red trip light on the SASS module indicates that: a. the channel has lost powe I
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b. the channel has NNI-Y power being supplied to i c. the channel has swapped to the alternate instrumen I d. MUV-31, Pressurizer Level Control, has transferred to manua ! QUESTION: 072 (1.00) The Shift Supervisor has been notified by one of the Fossil Units of a large, chlorine gas release. The Crystal River 3 Control Room has become uninhabitabl The Control Room Operators shall: a. notify plant personnel and evacuate the Control Roo b. trip the reactor and evacuate the Control Roo c. commence a plant shutdown and don air pack d. notify plant personnel and don air pack ' b '
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b ! QUESTION: 073 (1.00)- ,
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During a " Shutdown From Outside the Control Room," the Chief Nuclea Operator has locally opened CRD Breakers A, B, C,.and i
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The Chief Nuclear Operator must also verify: ; Source Range Nuclear Instrumentation on scale and DECREASIN I CRD groups 1 through 7 are fully inserte l boron concentration is INCREASIN , t I a negative 1/3 dpm start-up rate on the Source Range Nuclear Instrumentatio ;
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.C"ESTION: 074 (1.00)
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Given the following conditions: [
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A plant cooldown has been complete The Reactor Coolant System is at the refueling boron : concentratio !
. :What requirements must be met prior to opening both doors of Reactorf !
Building personnel and equipment hatches? ,
.The requirements in: 'the ODCM, "Off-site Dose Calculation Manual." AP-880, " Fire Protection."
- AI-502, "Defueled Plant Operations." OP-409, " Plant Ventilation Systems."
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Given the.following conditions:
-- 'Incore'temperatore indicate greater than 20 degrees superheated condition HPI is in operatio MUP-1B and -1C are on a single BWST suction lin BWST Level is 20 fee Select the suction source water supply for the HPI pump HPI suction is aligned to the: j Make-up Tan Borated Water Storage Tank (BWST). LPI system discharg Reactor Building sum .i QUESTION: 076 (1.00) :l Given the following conditions: --
Incore temperatures indicate greater than 20 degrees superhea Temperature of the cladding indicates within the " Region l IV" are :
Choose the requirements for Reactor Coolant Pump-(RCP) operatio ! Start ONE RCP per loop when all RCP start permissives are' met, Start ALL RCPs when all RCP start permissives are me .I Start ONE RCP per loop by bypassing RCP start permissive Start ALL RCPs by bypassing RCP start permissive I l
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SENIOR REACTOR OPERATOR Page 51 . . QUESTION: 077 (1.00) Given the following conditions:
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A reactor trip has occurred from 100% powe The "A" OTSG Main Feedwater Block valve, FWV-30, has stuck ope The Control Room operators have tripped both ain Feedwater pump Choose the Emergency Feedwater pump (EFW) start signal for the above condition Both EFW pumps start on a " Loss of Main Feedwater pumps." The steam driven EFW pump, EFP-2, is manually starte ' The electric driven EFW pump, EFP-1 is manually started, Both EFW pumps start on EFIC low level in the OTSG QUESTION: 078 (1.00) Given the following conditions:
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A reactor trip has occurred from 100% powe OTSG "A" and "B" pressures indicate 1030 psi Choose how OTSG pressure is being controlle OTSG pressure is being controlled by: a. the Turbine Bypass Valves (TBVs). the Main Steam Safetie the Turbine Bypass Valves (TBVs) and the Atmospheric Dump ; Valves (ADVs). l I the Atmospheric Dump Valves (ADVs) and the Main Steam Safetie ' O
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SENIOR REACTOR OPERATOR Page 52 . . QUESTION: 079 (1.00) Given the following conditions:
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A small break LOCA has occurre The reactor has been trippe Make-Up pumps are runnin Subcooling Margin by incore temperatures 35 degrees RCS pressure 1600 psig
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Combined flow of Make-Up Pumps 1000 gpm Select the " rule" for HPI contro a. HPI must be throttled to prevent pump runou b. HPI may be throttled anytime adequate subcooling margin exists based on incore c. HPI must be throttled to prevent exceeding NDT limit d. HPI may be throttled because LPI flow is greater than 1000 , gpm/line for greater than 20 minute ._- . . - -- _
-(3 '#
g' -- SENIOR REACTOR OPERATOR Page 53 i . . ,
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1 QUESTION: 080 (1.00) -l
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Given the following conditions: f
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Reactor power is 75%. !
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Pressurizer (PZR) level is 95 inches and DECREASIN !
-- An OTSG tube leak of 200 gpm is indicated in "A" OTS l t
Choose the REQUIRED operator actio : I Restore PZR level by opening the BWST suction valves, starting l the second make-up pump, and opening additional HPI valves, t i Close the Block Orifice Bypass valve (MU-51) and adjust the ;
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Pressurizer Level Make-up valve (MU-31) setpoint to maintain ; PZR leve j Restore PZR level by opening the BWST suction valves to supply- t
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make-up to the Make-up Tank and Pressurizer Level Control' valve (MU-31). ,
! Close the Letdown valve (MUV-49) and trip the reactor by j manually depressing the Reactor Trip pushbutton, j
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. SENIOR REACTOR OPERATOR Page 54 4 .
' QUESTION: 081 (1.00) Given the following conditions:
--
A Loss of Off-site Power has occurred coincident with a RCS leak of 20 gp The reactor trip was assumed to occur due to the Loss of Off-site Powe T-hot temperature is increasing towards saturation temperatur T-cold temperature has decreased below Tsat in the OTSG Select the cause for these indication The leak is located on the RCS Thot le The leak is located on the RCS Tcold le A loss of Natural Circulation has occurre Inadequate Core Cooling has develope QUESTION: 082 (1.00) Given the followir conditions:
--
Reactor Power is 100%.
--
RCS Pressure is decreasin Pressurizer Level is 220 inche i
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Pressurizer Temperature is decreasin l
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Make-up Tank level is constan i J Chocse the cause for these indication l a. A Pressurizer steam leak is occurrin b. A RCS leak in the letdown system is occurrin l c. The Pressurizer (PZR) level is below the heater cutoff leve d. The Pressurizer Spray Valve, RCV-14, is stuck ope _ .- . _ _ ._ O O : 4ENIOR REACTOR OPERATOR Page 55 ; . . ,
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QUESTION: 083 '(1.00) i Given the following conditions: ! i
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Source Range Nuclear Instrumentation counts read 20 counts l per second (cps). j
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Safety Rod Groups are at their "out-limit."
-- Group 5 rods are being withdraw ;
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Start-up rate and Source Range counts begin to increase- l rapidl Outward rod motion is stopped but source range counts continue to j
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increase and start-up rate is constant and positiv Select the cause for these indication l
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a. An inadvertent boration is in progres j b. The reactor is critica > c. Both Group 5 and 6 rods were withdrawn simultaneously because j of the overlap regio i
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d. Tave has increased in temperature during rod withdrawa 'i r QUESTION: 084 (1.00) j1 While performing a reactor start-up, the overlap between the Source and -! Intermediate Range Nuclear Instruments was 4 decade ) I Choose the cause for this amount of overla a. The Intermediate Range compensating voltage is set to LO .: b. The Source Range high voltage is set to HIG c. The Axial Power Shaping Rods (APSRs) are " Shadowing" the ! detector . d. A high boron concentration has caused the reactor to be "Over- l moderated." j f I i
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. SENIOR REACTOR OPERATOR e
Page 56 . . QUESTION: 085 (1.00) Given the following conditions:
--
A Main Steam line isolation has occurred at 600 psig on "B" OTSG and isolated a steam lea The resulting reactor trip and main steam line isolation has caused a OTSG tube rupture to occur in "B" OTS "B" OTSG steam pressure has recovered and is being maintained at 700 psi Choose the expected radiation monitor indication Condenser exhaust monitor, RMA-12, is increasin "B" OTSG Atmospheric Dump Valve (ADV) radiation monitor, RMG-28, is increasin "B" Main Steam Line radiation monitor, RMG-26, is increasin "A" Main Steam Line radiation monitor, RMG-27, is increasin QUESTION: 086 (1.00) Why is it desirable to steam the affected OTSG during a tube rupture, within TRACC limits, instead of isolating the affected OTS Steaming both OTSGs: reduces off-site doses significantl , i minimizes the duration of the cooldow l ensures that natural circulation will not be require l minimizes the amount of contamination in the main condense l l
'" . SENIOR REACTOR OPERATOR Page 57
+ . QUESTION: 087 (1.00) Given the following condition:
--
An OTSG tube rupture has occurred in ONE OTS The RCS leak rate is approximately 200 gp A Reactor trip has occurre Choose the parameter (s) used to identify the affected OTS The affected OTSG may be identified using: the condenser off-gas monitor, RMA-1 selected main steam header pressur RCS loop Tc temperature OTSG feed flows and OTSG level QUESTION: 088 (1.00) Given the following conditions:
--
A reactor trip has tccurred from 100% powe "B" OTSG had a stuck open steam safety which has rescate "B" OTSG level indicates 5 inches by EFIC low range indicatio Select the PREFERRED feedwater source and reason for using it to recover the "B" OTS a. Emergency Feedwater is preferred because recovery with Main Feedwater has not been analyze b. Emergency Feedwater is preferred because it sprays directly on the OTSG tube c. Main Feedwater is preferred because it is preheated in the downcomer prior to reaching the OTSG tube d. Auxiliary Feedwater is preferred because it is preheated in the downcomer prior to reaching the OTSG tube '
. SENIOR REACTOR OPERATOR Page 58 . .
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QUESTION: 089 (1.00) : , i l !
.Given the following condition: --
A bulb in the ICS delta Tc station was being replaced when l an-electrical short occurre !
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Both Main Feedwater' Block, Low Load. Block, and Start-up valves have shu .
--
The reactor has tripped on High RCS pressure. . l Choose the REQUIRED method of feed and level control for the OTSGs1With ! NO: operator actio Main Feedwater is controlling at low level limit ' Auxiliary Feedwater is controlling at low level limit l l Emergency Feedwater is controlling at 50% in the operating rang Emergency Feedwater is controlling at EFIC low level limit ; i i
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j r i QUESTION: _090 (1.00) !
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t-Given the following conditions: { i
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The reactor is in MODE l
' --
A containment purge is in progres Personnel are.in the Reactor Building doing a walkdown of the Reactor Coolant Syste : A HIGH FLOW alarm has been received by RM-A1, the containment purge gas l
: monito ;
Select the REQUIRED Control Room operator actio l i Isolate the Containment purge supply and exhaust valve j i
' Isolate the Reactor Building vent valves, LRV-70, 71, 72, and 7 j Stop the containment purge supply and exhaust fan f Sound the Reactor Building evacuation alarm and shut and-lock the Reactor Building Personnel Hatc j i
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Page 60
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QUESTION: 091 (1.00) Given the following conditions:
--
RCS pressure is 300 psig
--
Reactor Building pressure indicates 30 psig
--
No ES actuations have occurred
--
Health Physics reports detection of a severe containment radiation leak directly to the environment through a crack surrounding a penetratio Which of the following actions would RAPIDLY reduce the radiation leakage to the environment? Shift the Decay Heat Removal suction to the Emergency Sum Perform a Rapid reactor cooldown and depressurizatio Start two reactor building coolers and fan Initiate Reactor Building Spray with two pumps operatin l
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.SEMIOR REACTOR OPERATOR
. . QUESTION: 092 (1.00) Given the following conditions:
--
Instrument Air pressure indicates 80 psi The HPs have notified the Control Room that NO individuals are on breathing ai NSCCCS valve, SWV-110, " Return from the CRDMs," has failed close The Reactor Building sump level is slowly increasin Select the REQUIRED control room operator actio Trip the reactor and concurrently perform EOP-02, " Vital Systems Status Verification." Perform a controlled shutdown at the maximum safe rat Trip the reactor coolant pumps and perform EOP-09, " Natural Circulation Cooldown." Sample the Resetor Building sump to identify type of leakag QUESTION: 093 (1.00) Given the following conditions:
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Reactor Building sump level is increasin Pressurizer level is constant at 220 inche Make-up Tank level is decreasin Select the cause for these indication A small Main Steam leak inside the Reactor Buildin A small RCS leak on the letdown system in the Auxiliary Buildin A RCS leak within the capacity of normal make-u A leaking Pressurizer Code Safet
. SENIOR REACTOR OPERATOR Page 62
. . QUESTION: 094 (1.00) Given the following conditions:
--
Refueling operations are in progress inside the Reactor Buildin The Main Fuel Handling Bridge Operator has lifted a fuel assembly from the core into the mas RM-A1, Reactor Building Purge Duct radiation monitor has alarme Choose the REQUIRED actions to be taken by the Refueling Operator The Refueling Operators shall: lower the fuel assembly into the core and evacuate the are immediately evacuate the are nove the fuel assembly to the deep end of the pool and contact the Control Roo move the fuel assembly to the deep end of the pool and evacuate the Reactor Buildin l i
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. . SENIOR REACTOR OPERATOR Page 63
. . QUESTION: 095 (1.00) Given the following conditions:
--
A Loss of Off-site Power has occurre The running Make-up pump is NOT the ES selected pum Natural Circulation has been established and the plant is stable in Mode What is REQUIRED to re-establish Make-up and Purification and what precaution (s) should be taken with the REQUIRED actio The Make-up pump must be restarted and RCP seal injection must be closed and slowly re-establishe Re-establish letdown by bypassing the High temperature interlock and slowly increase flow back to the desired amoun Re-establish RCP seal controlled bleed-off and ensure the proper RCP seal staging is observe Place the Pressurizer Control Valve, MU-31, in automatic and ensure the Pressurizer Spray Valve, RCV-14, is shu . SENIOR REACTOR OPERATOR e'" Page 64 . . QUESTION: 096 (1.00) Given the following conditions:
--
A plant cooldown is in progres The Decay Heat System is in operatio The RCP(s) are ready to be secure What is the maximum step temperature change allowed when RCPs are secured and what temperatures are used to verify the step change? The maximum step change is: degrees F. using; (Tcold prior to stopping all RCPs) minus (Core Exit Thermocouples after the RCPs are off.) degrees F. using; (Tcold prior to stopping all RCPs) minus (Decay Heat Cooler Outlet temperature with RCPs off.) degrees F. using; (Thot prior to stopping all RCPs) minus (Core Exit Thermocouples after the RCPs are off.) degrees F. using; (Thot prior to stopping all RCPs) minus (Tcold after all RCPs are off.)
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. SENIOR REACTOR OPERATOR
. . QUESTION: 097 (1.00) A plant shutdown is in progress in "THREE Reactor Coolant Pump (RCP) Operation."
When the single RCP loop OTSG reaches low level limits, what will happen to Core delta T and the ICS feed flow circuitry? A Core delta Th will develop and both Main Feedwater Block valves will shut and Main Feedwater pump control circuitry will shift from speed control to " delta P" contro A Core delta Tc will develop and RCS flow input and Feedwater to Reactor Crosslimits to ICS will be disable A Core delta Th will develop and 2/3's of the feedwater flow error will be applied to the OTSG NOT on low level limit A Core delta Tc will develop and all feedwater flow error will be applied to the OTSG NOT on low level limit QUESTION: 098 (1.00) What is the BASIS for tripping all RCPs within 2 minutes when there is a Loss of Subcooling Margin? This keeps the RCS void fraction less than 70% such that a loss of RCPs would not uncover the core and peak cladding temperatures will not be exceede Securing the Reactor Coolant Pumps maintains electrical loading at a minimum, thereby allowing excess electrical power for use during the emergenc This allows the RCS to separate into a steam and water mixture which allows only high energy steam to escape from the brea With the RCPs secured, the delta p across the leak is reduced which prevents the leak size from increasin I
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f QUESTION: 099 (1.00) The Reactor power is 60%. There is a small RCS leak- (0.20 gpm) inside -, the Reactor Buildin Select _the following indicators.which indicate ! the RCS leak inside the Reactor Buildin Tave. constant -i RCS pressure' decreasing , Reactor Building sump level is constan ! Tave decreasing RCS pressure decreasing ; _ Reactor Building pressure is constan , t Tave constant l RCS pressure constant 't Reactor Building Fan "High" condensate alarms are illuminate l Tave decreasing RCS pressure decreasing Reactor Building temperature is constan ,
-l QUESTION: 100 (1.00) >
.
Given the following Conditions: '
--
A Natural Circulation cocidown is in progres Pressurizer level is increasin RCS pressure is decreasin ! Identify the cause for these plant condition , Normal plant cooldown condition . Reactor vessel head bubble formatio ; T-hot vents are ope Pressurizer spray valve, RCV-14, is ope ! i
; (********** END OF EXAMINATION **********)
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ANSWER: 001 (1.00) REFERENCE: Conduct of Operations, AI-500, Section 4.3.1.2, page 4 Lesson Plan, ROT-5-3 [4.1/3.9] 194001A102 ..(KA's) ANSWER: 002 (1.00) = REFERENCE: 10CFR20.101 ROT-5-43, LO B [2.8/3.4] l l e.
. $94001K103 ..(KA's)
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ANSWER: 003 (1.00) , l i
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'. SENIOR REACTOR OPERATOR e* Page 68
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REFERENCE: AI-412, " Verbal Communication Guideline."
LP, ROT-4-92, LO G [3.0/3.2] 194001A104 ..(KA's) ANSWER: 004 (1.00) REFERENCE: AI-500, " Conduct of Operations," page 9 & 10, steps 4.1.1.2 and 4.1. LP, ROT-5-38, LO B [2.5/3.4] 194001A103 ..(KA's) ANSWER: 005 (1.00) REFERENCE: AI-412, " Verbal Communication Guidelines," page 6, step 4. ROT-5-3 [3.6/3.8] 194001A105 ..(KA's)
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[_ ~. SENIOR' REACTOR OPERATOR O O Pago.69 , . . .
ANSWER: 006 (1.00). ; L l
' REFERENCE: '
EM-103, " Operation and Staffing of the CR-3 Control Room During Emergency Classifications," pago 2, Section 3. LP, ROT-5-34 l7 (3.1/4.4) l 194001A116 ..(KA's) ANSWER: 007 (1.00) REFERENCE: l EM-202, " Duties of the Emergency Coordinator," page 53, enclosure 4.
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-LP,_ ROT-5-34 [3.1/4.4]
194001A116 ..(KA's) ANSWER: 008 (1.00) b.
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____ _ _ _ _ _ _ _ _ _ _ _ - -__ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
.SENIGR REACTOR OPERATOR' Page 70
.o . REFERENCE: AI-1600, "Alara Program Manual," page 9, Section 4. LP, ROT-5-38
[3.3/3.5]
194001K104 ..(KA's) ANSWER: 009 (1.00) REFERENCE: CP-115, " Nuclear Plant Tags and Tagging Orders," page 13, Step 4. LP, ROT-5-40
[3.7/4.1]
194001K102 ..(KA's) ANSWER: 010 (1.00) REFERENCE: CP-115, Nuclear Plant Tags and Tagging Orders," page 23, step 4.11. LP, ROT-5-40
[3.7/4.1]
194001K102 ..(KA's)
- oSENIOR REACTOR OPERATOR Page 71
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ANSWER: 011 (1.00) REFERENCE: HPP-300, " Administrative Dose Limits and Dose Guidelines," page 4, step 3.4. LP, ROT-5-43
[3.3/3.5]
194001K104 ..(KA's) ANSWER: 012 (1.00) REFERENCE: AI-500, " Conduct of Operations," page 10, step 4.1. LP, ROT-5-38
[3.6/3.7]
194001K101 ..(KA's) ANSWER: 013 (1.00) ""
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SENIOR REACTOR OPERATOR Page 72 . . REFERENCE: CP-115, " Nuclear Plant Tags and Tagging Orders," page 12, Step 4. LP, ROT-5-40
[3.6/3.7]
194001K107 ..(KA's) ANSWER: 014 (1.00) REFERENCE: AI-400D, " Temporary Procedure Changes," page 7, Enclosure 1, step 1 LP, ROT-5-77
[3.1/4.1]
194001A112 ..(KA's) ANSWER: 015 (1.00) REFERENCE: AI-500, " Conduct of Operations," page 30, step 4.2.2 LP, ROT-5-38
[2.6/3.1]
194001A108 ..(KA's)
SENIOR REACTOR OPERATOR Paga 73 . . ANSWER: 016 (1.00) REFERENCE: AI-500, " Conduct of Operations," page 16, step 4.2. LP, ROT-5-38
[2.8/4.1]
194001A111 ..(KA's) ANSWER: 017 (1.00) REFERENCE: AI-500, " Conduct of Operations," page 17, step 4. LP, ROT-5-38
[2.9/3.9]
194001A110 ..(KA's) ANSWER: 018 (1.00) . .
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J.SENIGR REACTOR OPERATOR O O Page 74
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I REFERENCE * ROT-4-14, Integrated Control System, page 2 .LO?B . r
.[4.1/4.4]
-f 001000A106 ..(KA's) :
i i ANSWER: 019- (1.00) >
' REFERENCE:
'
OP-210, step 3. LP ROT-5-02, page 24
.i [4.2/4.3] ,
001000K518 ..(KA's) l, ANSWER: '020 (1.00) 1 REFERENCE: ; ROT-4-28, Control Rod Drive (electrical), Figure 3 ROT-4-28, LO > i
[4.5/4.4]
001000K105 ..(KA's)
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Page 75 . SENIOR REACTOR OPERATOR 4 . ANSWER: 021 (1.00) REFERENCE: OP-210, " Reactor Startup," page 4, step 3.2.5, and page 23, step ROT-5-02, page 25, F-4
[3.7/4.0]
001000K513 ..(KA's) ANSWER: 022 (1.00) REFERENCE: OP-302, "RC Pump Operation," Enclosure 1 and Check with facilit [3.5/3.9) i
.)
w a w ' , 003000A201 ..(KA's) ANSWER: 023 (1.00) REFERENCE:
[3.6/3.7]
f i i
a_ O [ h04000K520 ..(KA's)
. SENIOR REACTOR OPERATOR Page 76 . . ANSWER: 024 (1.00) REFERENCE:
[3.8/3.9]
004000A401 ..(KA's) ANSh'E R : 025 (1.00)
*
REFERENCE: ROT-4-14, " Engineered Safeguards Actuation System," Tables I & I LO B [4.2/4.4] 013000K106 ..(KA's) ANSWER: 026 (1.00) , i
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Page 77 4 g REFERENCE: OP-202, " Plant Heatup," page 59, step 4.5.13 ROT-5-02, page 23, LO [3.9/4.0)
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013000G013 ..(KA's) ANSWER: 027 (1.00) REFERENCE: ROT-5-7, " Heat Balance Calculations," page 1, step 3. LO B4
[2.6/3.1]
015000K504 ..(KA's) ANSWER: 028 (1.00) REFERENCE: ROT-3-03, " Natural Circulation," page 5, step LO B4
[3.4/3.7]
017020K401 ..(KA's)
_ - . . i
. SENIOR REACTOR OPERATOR O O Page 78 i
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i ANSWER: 029 (1.00)
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REFERENCE: i
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ROT-4-13, " Engineered Safeguards Actuation System," page 11,-Step 1. LO C6
-[3.3/3.C] ,
022000G007 ..(KA's) l
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ANSWER: 030 (1.00) i .!
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REFERENCE: ' ROT-4-15, " Emergency Feedwater and EFIC," page 5 ' B3
[3.7/3.7) .
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061000K202 ..(KA's) {
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ANSWER: 031 (1.00) i
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. SENIOR REACTOR OPERATOR Page 79 .j
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REFERENCE: i Emergency Operating Procedure, "EOP Entry Conditions," page 1 ROT-5-96,-page 2, LO B '[4.4/4.6]
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061000K301 ..(KA's) ;
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. ANSWER: 032 (1.00) ;
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' REFERENCE: ;
' Tech Specs, 3.9.9, page 3/4 9-9, " Containment Purge." . LO ROT-5-1, B10 '
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[3.6/3.8)
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071000A302 ..(KA's) ANSWER: 033 (1.00) ! f l l REFERENCE:
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Check With Facilit !
[3.4/3.6) )
:
072000A101 ..(KA's) .j
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~ ANSWER: 034 (1.00) )
, . -
.SENICR REACTOR OPERATOR ' Paga 80 . . REFERENCE: ROT-4-13, " Engineered Safeguards Actuation System," page 9, Section 1. LO B [3.6/3.7) 006020A303 ..(KA's) ANSWER: 035 (1.00) REFERENCE: ROT-4-60, " Reactor Coolant Systen," page 23 LO B4
[3.6/3.5)
010000A302 ..(KA's) ANSWER: 036 (1.00) REFERENCE: ROT-4-60, " Reactor Coolant System," page 24 LO B6
[4.0/3.8)
010000A403 ..(KA's)
. SENIOR REACTOR OPERATOR a'" Page 81 . . ANSWER: 037 (1.00)
- REFERENCE:
ROT-4-60, " Reactor Coolant System," page 2 LO B8
[3.3/3.7]
011000K401 ..(KA's) ANSWER: 038 (1.00) REFERENCE: RCT-4-60, " Reactor Coolant System," Check with facilit [3.1/3.1] 011000K604 ..(KA's) ANSWER: 039 (1.00) . SENIOR REACTOR OPERATOR e'" Page 82 . . REFERENCE: ROT-4-12, " Reactor Protection System," page 2 LO B [3.1/3.1] 012000K603 ..(KA'a) ANSWER: 040 (1.06) REFERENCE: ROT-4-12, " Reactor Protection System," LO B2
[3.1/3.5]
012000K603 ..(KA's) ANSWER: 041 (1.00) REFERENCE: ROT-4-28, " Control Rod Drive, Electrical," page 4 LO B2 [3.4/3.2] 014000A402 ..(KA's)
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1 . SENIOR REACTOR OPERATOR Page 83.-
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' ANSWER:. 042 -(1.00)
.:
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I ' REFERENCE: ROT-4-28, " Control' Rod Drive, Electrical," page 4 LO B2 [3.1/3.3]
-014000K405 ..(KA's) ,
r ANSWER: 043 (1.00) ;
, ;
t REFERENCE: , ROT-4-9, "Non-Nuclear Instrumentation," page 40 LO B ,
' [3.4/3.4]
i 016000K101 ..(KA's) Ei
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ANSWER: 044 (1.00) ,
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. SENIOR REACTOR OPERATOR Page 84 . . REFERENCE: ROT-4-9, "Non-Nuclear Instrumentation," page 5 LO B14 (3.2/3.3] 016000G004 ..(KA's) ANSWER: 045 (1.00) REFERENCE: ROT-4-29, " Spent Fuel Storage and Cooling System," page 9 LO B [3.0/3.3] 033000K303 ..(KA's) ANSWER: 046 (1.00) REFERENCE: EOP-11, " Loss of Decay Heat Removal," page 7, step LO ROT-4-32, "OTSG," B6 (3.4/3.6] 035000G001 ..(KA's) i
. SENIOR REACTOR OPERATOR Pagn 85 . . ANSWER: 047 (1.00) REFERENCE: ROT-4-91, "120 VAC Vital," page 1 LO F2
[3.1/3.5]
062000K410 ..(KA's) ANSWER: 048 (1.00) REFERENCE: OP-700E, "125/250 VDC Distribution Fanels," page 4 LO C:leck with facilit [3.7/4.1] 063000K301 ..(KA's) ANSWER: 049 (1.00) . SENIOR REACTOR OPERATOR Page 86
. . ., ' REFERENCE:
STS'3/4. 'LO ROT-4-54, " Decay Heat Removal," Bil
[3.2/3.6]
005000K307 ..(KA's) ANSWER: 050 (1.00)
' t REFERENCE: f
:
ROT-4-54, " Decay Heat Removal," page 1 ! LO B13 tj,
[3.5/3.6] ;
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005000G001 ..(KA's)
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ANSWER: 051 (1.00) ! i !
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REFERENCE: ~!
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ROT-4-60, " Reactor Coolant System." ' LO B3/B6 l
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[3.9/4.2] 1
:l 007000A201 ..(KA's)
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; ANSWER: 052 (1.00) '! ;
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~ REFERENCEi ROT-5-84, " Loss of Instrument Air."
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Lo Check with facilit :
-! [3.1/3.1] , '078000G015 ..(KA's) .'
di ANSWER: 053 (1.00) l l l REFERENCE: l -
' ROT-5-67, "AP-525 Continuous Control Rod Motion." 'I [3.2/3.6] [ ;
000001K122 ..(KA's)
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ANSWER: 054 (1.00) ,
.I-REFERENCE: 3 i
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[4.1/4.4]
000011K101 ..(KA's)
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; ANSWER: 063 (1. 00)
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000029K301 ..(KA's) . l'
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. SENIOR REACTOR OPERATOR PCga 92 . . ANSWER: 065 (1.00) REFERENCE: ROT-4-12, " Reactor Protective System," page 43 LO C [3.8/4.0] 000029G007 ..(KA's) ANSWER: 066 (1.00) REFERENCE: ROT-4-15, " Emergency Feedwater and EFIC," page 27 LO BIS
[4.6/4.6]
000040A101 ..(KA's) ANSWER: 067 (1.00) g g
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[4.0/4.1]
000040A106 ..(KA's) ANSWER: 068 (1.00) } REFERENCE: ROT-5-100, "EOP Tab 12, Station Blackout," page 6 LO B3
[4.1/4.4]
000055K102 ..(KA's) ANSWER: 069 (1.00) REFERENCE: EOP-01, " Entry Conditions."
LO ROT-5-96, "EOP TABS ... " B [4.1/4.1] 000055G011 ..(KA's)
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[4.4/4.6]
000055A202 ..(KA's) ANSWER: 071 (1.00) REFERENCE: ROT-5-81, " Loss of NNI-X," page 5 & LO CWF
[3.2/3.6]
000057A214 ..(KA's) ANSWER: 072 (1.00) ; J
. SENIOR REACTOR OPERATOR Paga 95 ;
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000068G001 ..(KA's) '
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.. . ANSWER: 075 (1.00) REFERENCE: EOP-07, " Inadequate Core Cooling," page LO ROT-3-25, B3
[3.9/4.1]
000074K204 ..(KA's) ANSWER: 076 (1.00) ] REFERENCE: EOP-07, " Inadequate Core Cooling," page 1 LO ROT-3-25, B3
[3.6/3.9]
000074A106 ..(KA's) ANSWER: 077 (1.00) L h'" b
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REFERENCE: ROT-4-15, " Emergency Feedwater and EFIC," page 1 LO B8
[4.0/4.6]
000007K301 ..(KA's) ANSWER: 078 (1.00) REFERENCE: EOP-02, " Vital System Status Verification," page 13 LO ROT-5-96, "EOP Tabs...," B3
[3.7/3.7]
000007A110 ..(KA's) ANSWER: 079 (1.00) REFERENCE: EOP-13, "HPI Control," Rule #2 ROT-5-96, LO B5
[4.4/4.4]
000009A113 ..(KA's)
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ANSWER: 081 (1.00) c.
l 'I REFERENCE: ROT-3-03, " Natural Circulation," page LO B4-(4.2/4.7] 000009K101 ..(KA's) ANSWER: 082 (1.00) d.
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[3.7/3.7]
000038A110 ..(KA's) ANSWER: 086 (1.00) REFERENCE: ROT-3-24, " Steam Generator Tube Rupture," page 16 LO B10
[4.2/4.5]
000038K306 ..(KA's) ANSWER: 087 (1.00) :
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[4.4/4.6]
000038A203 ..(KA's) ANSWER: 088 (1.00) REFERENCE: ROT-5-96, "EOP Tabs ..., " page 68 LO B5
[3.6/4.2)
000054K102 ..(KA's) ANSWER: 089 (1.00) REFERENCE: ROT-5-96, "EOP Tabs ..., " page 12 LO B5
[4.4/4.6]
000054K304 ..(KA's)
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ANSWER: 090 (1.00) REFERENCE: ROT-4-25, " Radiation Monitoring System," page 3 f LO F3A
[3.3/3.8]
000060G005 ..(KA's) l l ANSWER: 091 ( 1. O';) t REFERENCE: ROT-4-62 ) LO B [3.3/3.6] l < 000060G006 ..(KA's) ANSWER: 092 (1.00) I i 8.
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. SENIOR REACTOR OPERATOR Paga103 . . REFERENCE: AP-470, " Loss of Instrument Air," page 5 LO B1/B3
[3.7/3.9)
000065K308 ..(KA's) ANSWER: 093 (1.00) REFERENCE: ROT-4-52, "Make-up and Purification System," page 15 & 27 LO B14
[3.3/3.4]
000028A210 ..(KA's) ANSWER: 094 (1.00) REFERENCE: FP-203, "Defueling and Refueling Operations," page 8, step 3.2.16 LO ROT-4-26, F6
[3.4/3.9]
000056K202 ..(KA's) ANSWER: 095 (1.00) . SENIOR REACTOR OPERATOR a'"' A
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Page104 . . REFERENCE: ROT-4-52, "Make-up and Purification System," page 22 LO B15/B16
[3.8/3.9)
000056A203 ..(KA's) ANSWER: 096 (1.00) REFERENCE: OP-209, "Cooldown," page 37 Cneck with Facility
[3.7/3.8]
002000A103 ..(KA's) ANSWER: 097 (1.00) REFERENCE: ROT-4-14, " Integrated Control System," page 63 LO B1
[4.0/3.9]
035010A301 ..(KA's)
. SENIOR REACTOR OPERATOR a'"' Page105 . . ANSWER: 098 (1.00) REFERENCE: ROT-5-85, " Inadequate Subcooling Margin," page 4 LO A2
[3.7/4.2]
002000K514 ..(KA's) ANSWER: 099 (1.00) REFERENCE: ROT -3 -21, " Loss of Coolant Accidents," page 36 LO Terminal Objective #3
[4.3/4.4]
002000A201 ..(KA's) ANSWER: 100 (1.00)
.. [[ ' . SENIOR REACTOR OPERATOR Pago106 ) -.. . .
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001 a 024 c l40 002 b 025 )( 0. I 003 c 026 b 004 d 027 d 005 b 028 c-006 d 029 d
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007 c 030 c 008 b 031 a
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010 b 033 6 011 d 034 c 012 c 035 d 013 a 036 b 014 d 037 d 015 a 038 a
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016 c 039 b 017 b } gt 040 b 018 )(h 0 041 d 019 d 042 c 020 b 043 a 021 a 044 c ,
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022 b 045 d'
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Page a ' ANSWER KEY 046 a 068 d MULTIPLE CHOICE 069 b 047 b 070 a 048 c 071 c 049 b 072 d 050 c 073 b 051 c 074 a 052 a 075 c 053 b 076 d 054 c 077 a 055 a 078 c 056 d 079 b 057 a 080 d 058 b 081 c 059 c 082 d 060 c 083 b
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061 d 084 a
{3 062 a 085 ( )Io it 063 bp 086 b 064 d 087 d 065 a 088 a 066 c 089 d 067 b 090 c
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ANSWER KEY 091 d MULTIPLE CHOICE 092 a 093 c 094 b 095 a 096 b 097 d 098 a 099 c 100 b (********** END OF EXAMINATION **********) I
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TEST CROSS REFERENCE'
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SRO Exam PWR Reactor l
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Organized by Ques' tion Number f;
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QUESTION VALUE REFERENCE -f 001 1.00 9000001- l 002 1.00 9000002 003 1.00 9000003- ! 004 1.00 9000004 ., 005 1.00 9000005 ? 006 1.00 9000007 i 007 1.00 9000008 , 008 1.00 9000009- ! 009 1.00 9000010 010 1.00 9000011 011 1.00 9000012 1 012 1.00 9000013 013 1.00 9000014 ; 014 1.00 9000016 j 015 1.00 9000017
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't 016 1.00 9000018 017 1.00 9000019 -l 018 1.00 9000020 I 019 1.00 9000021 i 020 1.00 9000022 !
021 1.00 9000023 - 022 1.00 '9000024 023 1.00 9000025 024 1.00 9000026 i
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025 1.00 9000027 i 026 1.00 9000029 027 1.00 9000030 ;
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028 1.00 9000033 029 1.00 9000034 , 030 1.00 9000038 -2 031 1.00 9090039 032 1.00 9000041 i 033 1.00 9000043 034 1.00 9000048 ' 035 1.00 9000049 036 1.00 9000050 037 1.00 9000051 038_ 1.00 9000052 j 039 1.00 9000053 ' 040 1.00 9000054 ; 041 1.00 9000055 ; 042 1.00 9000056 043 1.00 9000057-044 1.00 9000058 - 045 1.00 9000059 ! 046 1.00 9000060 L'
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047 1.00 9000061 , 048 1.00 9000062 , 049 1.00 9000064 ; f
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SRO Exam PWR Reactor Organized by Question Number QUESTION VALUE REFERENCE 050 1.00 9000065 051 1.00 9000066 052 1.00 9000071 053 1.00 9000072 054 1.00 9000073 055 1.00 9000074 056 1.00 9000075 057 1.00 9000076 058 1.00 9000077 059 1.00 9000078 060 1.00 9000079 061 1.00 9000080 062 1.00 9000081 063 1.00 9000082 064 1.00 9000083 065 1.00 9000084 066 1.00 9000085 067 1.00 9000086 068 1.00 9000087 069 1.00 9000088 070 1.00 9000089 071 1.00 9000090 072 1.00 9000091 073 1.00 9000092 074 1.00 9000093 075 1.00 9000094 076 1.00 9000095 077 1.00 9000096 078 1.00 9000097 079 1.00 9000098 080 1.00 9000099 081 1.00 9000100 082 1.00 9000101 083 1.00 9000102 084 1.00 9000103 085 1.00 9000104 086 1.00 9000105 087 1.00 9000106 088 1.00 9000107 089 1.00 9000108 090 1.00 9000109 091 1.00 9000110 092 1.00 9000111 093 1.00 9000112 094 1.00 9000113 095 1.00 9000114 096 1.00 9000115 097 1.00 9000116 098 1.00 9000117
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L TEST CROSS REFERENCE Page 3 . . SRO Exam PWR Reactor Organized by Quo cion Number QUESTION VALUE REFERENCE 099 1.00 9000118 100 1.00 9000119 ___ _ 100.00 __ ___ 99 59 GuS IND _ _ 100.00
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b __) s TEST CROSS REFERENCE Page 4 . v SRO Exam PWR Reactor Organi zed by KA Group PLANT WIDE GENERICS QUESTION VALUE KA 001 1.00 194001A102 004 1.00 194001A103 003 1.00 194001A104 005 1.00 194001A105 015 1.00 194001A108 017 1.00 194001A110 016 1.00 194001A111 014 1.00 194001A112 006 1.00 194001A116 007 1.00 194001A116 012 1.00 194001K101 009 1.00 194001K102 010 1.00 194001K102 002 1.00 194001K103 011 1.00 194001K104 ' 008 1.00 194001K104 013 1.00 194001K107 ______ PWG Total 17.00 PLANT SYSTEMS Group I QUrsfION VALUE KA 018 1.00 001000A106 020 1.00 001000K105 021 1.00 001000K513 019 1.00 001000K518 022 1.00 003000A201 024 1.00 004000A401 023 1.00 004000K520 026 1.00 013000G013 025 1.00 013000K106 041 1.00 014000A402 042 1.00 014000K405 027 1.00 015000K504 028 1.00 017020K401 029 1.00 022000G007 030 1.00 061000K202 031 1.00 061000K301 048 1.00 063000K301 032 1.00 071000A302 033 1.00 072000A101 ______
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.s: TEST CROSS REFERENCE Paga. 5
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l 0rganized by KA Group , b PLANT SYSTEMS Group I- , QUESTION VALUE KA .i , PS-I Total 19.00 < Group II QUESTION VALUE KA i 096 1.00 002000A103 099 1.00 002000A201 100 1.00 002000G014 f 098 1.00 002000K514 I 034 1 00 006020A303 035 1.00 010000A302 036 1.00 010000A403 i 037 1.00 011000K401 038 1.00 011000K604 040 1.00 012000K603 039 1.00 012000K603 l 044 1.00 016000G004 043 1.00 016000K101 - 045 1.00 033000K303 '! 046 1.00 035000G001 097 1.00 035010A301 . 047 1.00 062000K410 ______
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PS-II Total 17.00 , Group III , QUESTION VALUE KA _
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050 1.00 005000G001 049 1.00 005000K307 051 1.00 007000A201
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052 1.00 078000G015
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______ PS-III Total 4.00 ______ ______ PS Total 40.00
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'# .4 TEST CROSS REFERENCE- Page 6-e e SRO Exam PWR Reactor organized by KA Group- ' EMERGENCY PLANT EVOLUTIONS Group I QUESTION VALUE KA 053 1.00 000001K122 054 1.00 000003A103 055 1.00 000003G009 056 1.00 000005A105 057 1.00 000011A115 058 1.00 000011A213 059 1.00 000011K101 060 1.00 000015A201 i'
061 1.00 000015G007 062 1.00 000026K303 064 1.00 000029A209 065 1.00 000029G007 063 1.00 000029K301 066 1.00 000040A101 067 1.00 000040A106 > 070 1.00 000055A202 069 1.00 -000055G011 068 1.00 000055K102 071 1.00 000057A214 072 1.00 000068G001 l
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073 1.00 000068K301 074 1.00 000069G007 076 1.00 000074A106
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075 1.00 000074K204 ______ EPE-I Total 24.00 , Group II QUESTION VALUE KA 078 1.00 000007A110 077 1.00 000007K301 079 1.00 000009A113 l 080 1.00 000009A206 -i 081 1.00 000009K101 082 1.00 000027A101 083 1.00 000032A202 084 1.00 000033A211 ; 085 1.00 000038A110 ; 087 1.00 000038A203 " 086 1.00 000038K306 ') 088 1.00 000054K102 l 089 1.00 000054K304 3 090 1.00 000060G005
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SRO Exam PWR Reactor Organized by KA Group , EMERGENCY PLANT EVOLUTIONS Group II QUESTION VALUE KA 091 1.00 000060G006 092 1.00 000065K308 ______ EPE-II Total 16.00 Group III QUESTION VALUE KA 093 1.00 000028A210 095 1.00 000056A203 094 1.00 000056K202 ______ EPE-III Total 3.00 ______ ______ EPE Total 43.00 ______ ______ ______ Test Total 100.00 ___ ;
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U. S. NUCLEAR REGULATORY COMMISSION SITE-SPECIFIC WRITTEN EXAMINATION APPLICANT INFORMATION Name: Region: II Date: 1993/11/15 Facility / Unit: Crystal River 3 License Level: R0 Reactor Type: BW INSTRUCTIONS Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. The passing grade requires a final grade of at least 80 percent. Examination papers will be picked up 4 hours after the examination start All work done on this examination is my ow I have neither given nor received ai Applicant's Signature RESULTS Examination Value 100.00 Points Applicant's Score Points Applicant's Grade Percent
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, REACTOR OPERATOR Page 2 ANSWER SHEET Multiple Choice (Circle or X your choice) If you change your answer, write your selection in the blan MULTIPLE CHOICE 023 a b c d 001 a b c d 024 a b c d 002 a b c d 025 a b c d 003 a b c d 026 a b c d 004 a b c d 027 a b c d 005 a b c d 028 a b c d 006 a b c d 029 a b c d , 007 a b c d 030 a b c d 008 a b c d 031 a b c d 009 a b c d 032 a b c d 010 a b c d 033 a b c d 011 a b c d 034 a b c d 012 a b c d 035 a b c 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 a b c d 043 a b c d 021 a b c d 044 a b c d 022 a b c d 045 a b c d
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, REACTOR' OPERATOR .Page. 3 A'N.S W E R ' SHEET Multiple Choice .(Circle or X your choice)'
If you change your answer, write your selection in the blan a b c d 068 a b c d MULTIPLE CHOICE 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 n
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, REACTOR OPERATOR Page 4 ANSWER SHEET Multiple-choice (circle or X your choice)
If you change your answer, write your selection in the blan a b c d MULTIPLE CHOICE 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 ( * * * * * * * * * * 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: Cheating on the examination will result in a denial of your application and could result in more severe penaltie . After you complete the examination, sign the statement on the cover sheet indicating that the work is your own and you have not received or given assistance in completing the examinatio . To pass the examination, you must achieve a grade of 80 percent or greate . The point value for each question is indicated in parentheses after the question numbe . There is a time limit of 4 hours for completing the examinatio . Use only black ink or dark pencil to ensure legible copie . Print your name in the blank provided on the examination cover sheet and the answer shee . Mark your answers on the answer sheet provided and do not leave any question blan . If the intent of a question is unclear, ask questions of the examiner onl . Restroom trips are permitted, but only one applicant at a time will be allowed to leave. Avoid all contact with anyone outside the examination room to eliminate even the appearance or possibility of cheatin . When you complete the examination, assemble a package including the examination questions, examination aids, and answer sheets and give it to the examiner or procto Remember to sign the statement on the examination cover shee . After you have turned in your examination, leave the examination area as defined by the examiner.
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QUESTION: 001 (1.00) Given the following conditions:
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Make-Up and Purification Pump, MUP-1A, quarterly surveillance is in progres RCP Seal Injection Flow Indicator, MU-27-FI, indicates 35 gpm and flow needs to be adjuste RCP Seal Injection Flow Control Valve, MUV-16, is Blue-tagge The Blue Tag instructs the operator NOT to adjust MUV-16 due to packing leakag Select the operator action required to complete the surveillanc Stop the surveillance and contact the Nuclear Shift Superviso Read the Blue Tag and continue with the surveillance, Send an operator to monitor MUV-16 and continue with the surveillanc Clear the Blue Tag, continue with the surveillance, and notify the Nuclear Shift Supervisor of action taken when the surveillance is complete da
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QUESTION: 002 (1.00) Given the following conditions:
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A 21 year old male radiation worke A current NRC Form 5 on file for the worke He has received 47.0 rem to the skin of the whole body this year (SDE-WB)
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His current Annual Total Effective Dose Equivalent (TEDE) dose is According to 10CFR20.101, what is the MAXIMUM amount of Total Effective Dose Equivalent (TEDE) he can receive for the remainder of the year? .5 re .0 re .0 re .0 re QUESTION: 003 (1.00) Which plant communication system is dedicated for fire reporting? PL-2 on the PAX System, Channel 5 on the hand held radio PL-1 on the PAX Syste Channels 13 and 14 on the hand held radio _
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QUESTION: 004 (1.00) The Nuclear Shift Supervisor (NSS) directs you, with face to face communication, to line up the Make-Up and Purification Pump, MUP-1B, for the annual surveillanc What communication requirement is NOT necessary for this exchange? The use of: repeat backs or paraphrasing the message, l i title /name identifier j i' confirmation from the NS I proper names and equipment number ) l l
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I QUESTION: 005 l Given the following conditions:
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A reactor trip occurred at 090 During shift turnover, the ONCOMING, evening shift, Nuclear Auxillary Operator (NAO), notices that the previous shift Auxiliary Operator logs have NOT been take Select the required action to be taken by the NA Perform a shift tour and: complete the logs for the previous shif complete the logs for the previous shift and your own shift at the same tim leave the previous shift logs blan notify the Shift Supervisor for a determination on what action to take for the blank log _ _ - _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ .
b --
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QUESTION: 006 (1.00) Given the following conditions:
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A small break loss of coolant accident (LOCA) is in progres The Nuclear Shift Supervisor has declared an Alert and announced it over the plant paging syste You are the On-Shift Nuclear Auxiliary Operato Where should you report to upon hearing the declaration of Alert? The Technical Support Center (TSC) The Operational Support Center (CSC) The Operations Conference Room The Control Room QUESTION: 007 (1.00) Given the following conditions:
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A General Emergency has been declare A radioactive release is occurring from the Auxiliary Buildin A Radiochemist has injured her leg and cannot exit the Auxiliary Buildin Choose the requirements for an emergency worker to assist the radiochemis The worker must be a male and have his dose limited to 75 rem whole bod The worker can be male or female and have his or her dose limited to 25 rem whole bod The worker must be a volunteer and have their dose limited to 75 rem whole bod The worker can be male or female and his or her exposure should not exceed quarterly exposure limit _ _ _ - . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ - _ _
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QUESTION: 008 (1.00) Given the following conditions:
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The plant has been in Mode The Letdown and Purification System has had extensive work performed on i A valve line-up on the Letdown and Purification System inside the Reactor Building is require Select your responsibilities as a radiation worker while performing the valve line-u Decline to do the job because the dose rates are to hig Perform the job quickly and safely to reduce the time spent in the radiation are Perform the job by verifying the valve positions using the previously completed valve line-u Perform the job from a distance by using a flashlight and direct line of sight to observe valve positio QUESTION: 009 (1.00) You are performing an independent verification (2nd checker) of a tagout on the High Pressure Injection System. The "First Person" clearing the tagout asks you to observe him repositioning a throttle valv Select the Independent verifier (2nd checker) respors This is appropriate because it is a throttle valv b. This is incorrect because it is NOT " Independent Verification."
c. This is appropriate because it is restoration of a non-safety system componen This is incorrect because Independent Verification is required to be a " hands on" performanc ;
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. i " EQUESTION: 010 (1.00) , i Breathing air is: lined.up to the Reactor Building for use by the j
. mechanics to replace a Reactor Coolant Pump-(RCP) Sea j
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What prevents lthe isolation of the breathing air supply valve? The Breathing Air Supply valve is: , a. red tagged ope , b. blue tagged ope , c. white tagged ope d. locked open with no ta ,
; -QUESTION: 011 (1.00)
e Given the following conditions:
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Refueling, operations are in progres New fuel is-in the transfer carriage and traveling from the spent fuel pool to the reactor buildin .;
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The SRO and RO monitoring refueling operations are on the c Main Fuel Handling. Bridg , A Nuclear Operator (NO) comes up to the bridge with'a tag out requiring
'an independent verification on a tag located a short distance awa Select the allowed action of the refueling operators -(RO & SRO) .
a. The Refueling RO verifies the tag since the Refueling SRO can NOT have any other concurrent dutie l b. The Refueling SRO verifies the tag since the Refueling RO is { NOT allowed to leave the Main Fuel Handling Bridg ,
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c. Neither the Refueling SRO.or RO verify the tag and inform the , NO that he must use another operator for verificatio l
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QUESTION: 012 (1.00) Given the following conditions:
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A reactor trip has occurre The trip was due to a fault on one of the Main Generator output breaker The Shift Supervisor has directed you to make up the tagout for the work on the Main Generator output breakers. What must be completed prior to hanging the in-plant tagout? A system clearance from the Dispatcher to the on duty Shift Supervisor must be in plac A grcunding device must be installed on the main generator side of the main generator output breaker The Emergency Diesel Generators must be tested and one left running and loaded on a 4160 bu The Main Generator must be purged of hydrogen and filled with nitrogen prior work on the output breaker QUESTION: 013 (1.00) There are RED flashing alarms on the "Cer^er Overhead CRT." The operator depresses the "ACK" ke What happens to these alarms? a. They are deleted from the scree I b. They come in solid with a green backgroun J c. They come in solid with a white backgroun d. They remain flashing with a white backgroun i l
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QUESTION: 014 (1.00) Given the following conditions:
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A Reactor Startup is in progress, ECP is 30% on Group Group 7 rods are being withdrawn and are at 28%.
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Rod 7-5 drops and its "in-limit" lamp is illuminate Start-up rate indicates NEGATIVE 0.10 dpm, counts are decreatin The Reactor Operator has stopped withdrawing control rod Choose the action required to recover the dropped ro Determine the cause for the dropped rod: ,
. and recover the dropped rod by withdrawing it to the grou and drive Group 7 rods to the in-limit, relatch all Group 7 rods, and continue with startu and continue the reactor startup until Group 7 rods are at 30%,
then recover the dropped ro and at a minimum, fully insert rod groups 5, 6, and 7 and reverify initial conditions of OP-21 ,
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i QUESTION: 015 (1.00)
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Given the following conditions:
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Reactor power is 100%.
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An approved, special Reactor Protection System (RPS) test is in progres During the test, " Electronic trips E & F" are actuated i simultantousl l
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Channel trips A, B, C, and D are normal (NOT TRIPPED).
i Select the expected Control Rod Drive (CRD) respons I All rods fully insert (Drop).
.! Regulating Rod Groups 5, 6, and 7 insert (Drop). Safety Rod Groups 1, 2, 3, and 4 insert (Drop). [ Regulating Rod Groups 5, 6, and 7 remain in position but cannot ! be move i
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QUESTION: 016 (1.00) ;
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Given the following conditions: !
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Reactor Coolant Pump 3B2 (RCP-1D) - Controlled Bleed Of f flow l has decrease ;
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RCP-1D Seal leakoff flow has increase l
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RCP Seal Injection Flow is constan j F
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Select the cause for the abnormal indication First stage RCP seal degradation is occurrin Third stage RCP seal degradation is occurrin ! RCS letdown has-been increase , I RCS pressure has decrease !
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I i I QUESTION: 017 .(1.00) j i Given the'following conditions: j
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Reactor power is 100% and stabl All ICS stations are in " automatic." 1
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Control rods have been moving INWARD over the past-24 } hour i'
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The RCS leak rate has DECREASED over the past 24 hour Tave and pressurizer level have been constan Select the cause for these change !
! Xenon has been increasing to reach equilibriu )
1 Inadvertent boration is occurrin , Inadvertent boron dilution is occurrin [ t Samarium has been decreasing to reach equilibrium, i l
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i QUESTION: 018 (1.00) d
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Given the following conditions:
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Reactor power is 90% and stable, j
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The Reactor Demand station is in " manual."
-- ICS is in " track." ;
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The remaining ICS stations are in " automatic." r
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A 3 ppm boron dilution is performe 'l
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30 minutes has elapsed since the dilution with NO ! additional operator actions take Which one of the following describes Tave's expected response? -j i Tave will: l
: vary and stabilize at a higher valu 'j decrease and stabilize at a lower valu f-l vary and return to the same valu j decrease and then increase to a higher valu lj
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QUESTION: 019 (1.00) Given the following conditions:
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Reactor Coolant System (RCS) pressure 1490 psi Reactor Building pressure 3.5 psi MUP-1A, Make-Up pump 1A Runnin MUP-1B Selected to B ES-416 MUP-1C In ES standb Select the mode of operation for the Make-Up Pump MUP-1A is running MUP-1B 1s running A ES i B ES MUP-1C is running $ A 8 I c g MUP-1A is running MUP-1B is running MUP-1C is off
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, MUP-1A is off MUP-1B is running __
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QUESTION: 020 (1.00) Concerning the Engineered Safeguards Actuation System (ESAS); How do the Relay and Channel Cabinets " operate" in order to send a " Trip Signal?" The Channel Cabinets de-energize to send a trip signa The Relay Cabinets energize to send a trip signa The Channel Cabinets de-energize to send a trip signa The Relay Cabinets de-energize to send a trip signa The Channel Cabinets energize to send a trip signa The Relay Cabinets energize to send a trip signa The Channel Cabinets energize to send a trip signa The Relay Cabinets channels de-energize to send a trip signa QUESTION: 021 (1.00) Given the following conditions:
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RCS pressure 1625 psig
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Reactor Building pressure 0.50 psig
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A plant heatup is in progres What must be done to the ESAS PRIOR to exceeding 1650 psig in the RCS? Place LPI in ES Standb Place HPI in ES Standb Open both Core Flood Octlet Valves and lock open the breaker Renove the shutdown bypass from all RPS channel r-O
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QUESTION: 022 (1.00) When NI-5 exceeds a trip setpoint and sends a trip signal, what occurs in the RPS Channels? Each CRDM breaker and electronic trip relay will have ONE set of contacts open, Only Channel "A" CRDM breaker will have one set of contacts ope Only Channel "A" CRDM breaker will have both sets of contacts ope Each CRDM breaker and electronic trip relay will have BOTH sets of contacts open, QUESTION: 023 (1.00) Given the following conditions:
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Reactor power is 1.0 E-10 amp Control Rods are being withdrawn to achieve 1.0 E-8 amps reactor powe The Control Room receives a report that the Atmospheric Dump Valves (ADVs) are leakin At the same instant in time as the dump valve report, the control rods cease outward movemen Choose the cause for the stoppage of outward rod motio The 2 dpm Startup Rate Rod interlock in the source range, The 3 dpm Startup Rate Rod interlock in the intermediate rang The " Point of Adding Heat" has been reached causing the 3 dpm Startup Rate Rod interlock to activat Neutron error is greater than 1% causing the Diamond Panel Control F*ction to shift to " Manual."
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QUESTION: 024 (1.00) Given the following conditions:
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A station blackout has occurre Natural Circulation has been establishe The plant is in " Hot Standby."
-- Subcooling Margin is 45 degrees The Shift Supervisor directs the Chief Nuclear Operator #1 (CNO 1) to verify the subcooling margin with the use of the Steam Table Select the instrumentation the CNO 1 will use for this verificatio RCS T-cold leg temperature and RCS narrow range pressor RCS T-hot leg temperature and RCS wide range pressur Core exit thermocouples and RCS wide range pressur Core exit thermocouples and OTSG saturation pressur QUESTION: 025 (1.00) The Reactor Building Cooling and Isolation System will limit the post accident ambient pressures to design va]ue What is the MINIMUM combination of Reactor Building (RB) Spray System Trains and Emergency Cooling Units AHF-1A, -1B, and -1C needed to meet this design criteria? ONE RB Spray Train and ONE Emergency Cooling Uni , b. NO RB Spray Trains and TWO Emergency Cooling Unit ONE RB Spray Train and NO Emergency Cooling Units, NO RB Spray Trains and THREE Emergency Cooling Unit ! l l l
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QUESTION: 026 (1.00) Given the following conditions:
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The Reactor has tripped from 100% powe All ICS stations were in " automatic" prior to the reactor tri Choose the signal that is reducing feedwater flo BTU limits Reactor liruited by feedwater crosslimit Unit Load Demand (ULD) signal of 20% per minute Rapid Feedwater Reduction (RFR) QUESTION: 027 (1.00) Given the following conditions:
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A power increase from 45% to 75% is in progres Both Main Feedwater pumps are operatin Each Main Feedwater pump is being controlled by its Main Feedwater Block valve " delta p."
-- Loop feedwater demands read 50%. Select the expected main feedwater respons Feedwater pump control will shift from " delta p" control to flow erro The Main Feedwater Block valves will pulse close The Low Load Control valves will release to control feedwate The Start-up Valves will freeze in positio * O
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. . QUESTION: 028 (1.00) Given the following conditions:
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A Loss of offsite power has occurre Subcooling margin was less than 30 degrees Emergency Feedwater (EFW) is running and feeding both Once Through Steam Generators (OTSGs).
Offsite power has been restored and subcooling margin is 50 degrees One Reactor Coolant Pump (RCP) per loop has been restarte Select the expected OTSG level respons OTSG level will decrease: from 95% to 65% level setpoin from 95% to 30 inch level setpoin from 65% to 30 inch level setpoin from 65% to 24 inch level setpoin QUESTION: 029 (1.00) Given the following conditions:
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A Loss of Offsite Power has occurre ESAS has actuated on 1500 psig RCS pressure decreasin What is the power supply to the motor driven Emergency Feedwater Pump? Emergency Diesel Generator 1B Unit Bus 3 Emergency Diesel Generator 1 Unit Bus 3 l
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, . . QUESTION: 030 (1.00) l Given the following conditions:
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A reactor trip has occurred from 100% powe Concurrent with the reactor trip, a loss of both Main and Emergency Feedwater has occurred.
l l Select the required EOP section to be entered for these condition EOP-02, " Vital System Status Verification." EOP-04, " Inadequate Heat Transfer." EOP-05, " Excessive Heat Transfer." EOP-07, " Inadequate Core Cooling."
QUESTION: 031 (1.00) Given the following conditions:
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A liquid radwaste release is in progress from the l Evaporator Condensate Storage Tank (ECST), WDT-10 l
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The release is via the " Preferred Release Path."
-- Raw Water Pumps 1 and 2A are runnin Circulating Water Pumps are runnin ! During the release, WDV-892 (ESCT discharge isolation valve) automatically close Choose the cause for the automatic closure of WDV-89 Radiation Monitor, RM-L2, indicates background count rat Both Circulating Water pumps have trippe Both Raw water pumps have trippe WD-101-FR (Flow Recorder) is NOT operable.
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t . .. QUESTION: 032 (1.00) Given the following conditions:
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Refueling operations are in progress inside the Reactor Buildin A continuous Reactor Building purge is in progres RM-Al is required to be removed from service for calibratio Select the required operator action when RM-Al is removed from Service, Isolate containment purg Take a grab air sample every 4 hours, and continue containment purg ! Bypass the RM-Al interlocks and place a temporary radiation monitor on the refueling dec Require all personnel inside the reactor building to wear respirator . QUESTION: 033 (1.00) ; Select the function, as required by Technical Specifications, for the ' Containment Iodine and Gaseous Monitor, RMA- I RMA-6 is: a, used for Reactor Building Purge permit ;
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' part of the RCS leak detection syste used for Fuel handling operation > part of the post accident sampling syste !
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. . QUESTION: 034 (1.00) Given the following conditions:
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A reactor trip has occurred from 100% powe " Radiation High Alarm,"is illuminated for Radiation Monitor RM-L RM-G6 and RM-G10, Make-up Tank and Pump Room radiation monitors indicate increasing radiation level The Make-up Pre-filter High " Delta p" Alarm is illuminate Select the cause for these indications, Failed fuel has occurred in the reactor cor The make-up demineralizers have faile There is a RCS leak in the make-up tank roo The reactor trip has caused a crud burs QUESTION: 035 (1.00) Given the following conditions:
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A Loss of Off-site power has occurre "A" OTSG has been isolated due to a steam lea Core exit thermocouples indicate increasing temperature > Select the method that will maintain adequate natural circulation flow, Increase the OTSG level in the "A" OTSG above the EFIC high range setpoin Increase the saturation temperature in the "B" OTS Decrease the saturation temperature in the "A" OTS Increase the OTSG level in the "B" OTSG above the EFIC high range setpoin t T
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QUESTION: 036 (1.00) Given the following conditions:
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Reactor power has been held at 90% for Xenon to approach equilibriu During the power increase to 100%, control rods withdraw more than anticipated and Group 7 reaches the "outlimit."
Choose the reason for the reactivity differenc Xenon concentration was less than expecte The boron dilution performed for power increase was too large, Power doppler was not accounted for in the power increase, Samarium concentration was less than expecte QUESTION: 037 (1.00) Given the following conditions:
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A reactor trip has occurre Subcooling Margin (SCM) is 28 degrees RCS pressure is 1700 psi Main Steam Isolation has occurred on "B" OTS The overcooling transient has been terminated at a RCS pressure of 1650 psi What SCM is required to be maintained? SCM is required to be maintained greater than: degrees F. subcooled, but less than 30 degrees degrees F. subcooled, but less than 50 degrees c. 50 degrees F. subcooled, but less than 70 degrees degrees F. subcoole . REACTOR OPERATOR .Page 27 .. . QUESTION: 038- (1.00) Given the following conditions:
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The Engineered Safeguards Actuation System (ESAS) has initiated on 4.0 psig Reactor Building pressur Main Steam Isolation has occurre The Reactor Building Isolation and Cooling (RBIC) System has actuate The CNO bypasses RBIC actuation to restore RCP service Reactor Building pressure decreases initially to 2.0 psig and then increases to 5.0 psi : Select the RBIC system respons The RBIC system: , t
. re-actuate i
! bypass lights go from solid to slow flas . remains bypassed because of the seal-in logi re-actuates e> cept f or the RCP service , 1
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s . ,) QUESTIONi 039 (1.00)
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Reactor Building pressure is 3.0 psig and increasing'. 'The" Shift . Supervisor,Lanticipating 4.0 psig building pressure, has you manuall , actuate the LPI system by depressing both pushbutton What mode of operation are the LPI pumps in? , The LPI pumps are: l
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a. running and on recirculation until RCS pressure lowers to 500
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psi b. running and flow is being injected into the RC ! c. idle (NOT running) , because the "HPI seal in" is NOT presen I d. idle (NOT running), until the LPI system valves realign,-and' then they' receive an auto-start signa >
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QUESTION: 040 (1.00)
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i Given the following conditions: ~l t
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RCS pressure is 2205 psig and increasin .. ;
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The Pressurizer Spray Control valve " green" indicating ; light has' extinguished and one amber light has illuminate ;
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A second amber light on the Pressurizer Spray Control 1 valve has illuminate ' Choose the cause for the illumination of the second amber ligh ! a. The Pressurizer Spray valve is 100% ope .; b. The-Pressurizer Spray Block valve is close c. The Pressurizer heater banks "A" and "B" are of ! d. The Pressurizer Spray valve is 40% ope l
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. . QUESTION: 041 (1.00) * Given the following conditions:
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The Pilot-Operated Relief Valve (PORV/RCV-10) is in
" automatic."
-- RCS pressure has increased to the PORV automatic open setpoin The PORV " red" and " amber" lights are illuminate Select the status of the PORV/RCV-10 in this mod The PORV: is 40% ope solenoid valve is energize is 100% ope is " interlocked" close : i
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QUESTION: 042 (1.00) Given the following conditions:
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A reactor trip from 75% power has occurre A malfunction in Rapid Feedwater Reduction (RFR) circuitry has caused main feedwater to overfee Pressurizer heater banks "D" and "E" have been placed in "on" but they did NOT energiz Select the cause for the lack of power to Pressurizer heater banks "D" and "E". An insurge into the pressurizer has caused RCS pressure to increase to 2155 psi The Pressurizer Spray valve "Open Interlock" has removed power from pressurizer heater bank The Pilot-Operated Relief Block Valve (PORV) is closed and is interlocked with the pressurizer heaters, An outsurge from the pressurizer has caused level to decrease below 40 inche * REACTOR OPERATOR e
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Pressurizer level and pressure control are in " automatic."
-- A small leak (0.20 gpm) develops in the pressurizer steam spac Select the expected Pressurizer and Make-up Tank level respons Pressurizer level remains constan Make-up Tank level decreases, Pressurizer level increase Make-up Tank level decrease Pressurizer level increase Make-up Tank level increase Pressurizer level decrease Make-up Tank level increases QUESTION: 044 (1.00) Which condition is used to indicate to RPS that the Main Turbine is tripped? The EHC oil pressure is less than or equal to 45 psi The Auto Stop oil pressure is less than or equal to 45 psi The Turbine Throttle valve LVDTs indicate close The Turbine Governor valve LVDTs indicate close ,
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QUESTION: 045 (1.00) Given.the following conditions:
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NI-5'has failed and Channel "A" Reactor Protection System (RPS) . is in the " Trip Condition. " ,
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The I&C technician must perform NI-6, Channel "B" RPS, ,
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testinc A." Minimum Degree of Redundancy" of ONE must be maintaine _ ! l What must be done with RPS Channels "A" and "B" in order to perform ! Channel "B" testing? i i Place Channel "B" RPS: l in " bypass" and Channel "A" RPS in " bypass." in " bypass" and leave Channel "A" RPS " tripped." ! t in " shutdown bypass" and leave Channel "A" RPS " tripped." in " Normal" and leave Channel "A" RPS " tripped."
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QUESTION: 046 (1.00) Given the following conditions:
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An electrical fault has caused erratic indication in Channel "A" RP The SASS monitors are NOT operationa RCS pressure is mimicking the Channel "A" RPS indication Select the cause for RCS pressure fluctuation The Non-Nuclear Instrumentation (NNI) control signal is selected to Channel "A" RP The Nuclear Instrumentation (NI) Channel 5 signal is supplying IC The RCS flow signal supplied to the Main Control Board (MCB) is selected to Channel "A" RP The RCS flow signal supplied to the ICS is selected to Channel
"A" RP QUESTION: 047 (1.00)
The selected Main Steam Header Pressure instrument is degrading (failing) from 895 psig to zero over 5 minute Choose the SASS response to this failur The SASS channel: trips and swaps control to the non-changing signa trips and the amber mismatch lamp is illuminate c. reverts to the " manual" mode and the amber mismatch lamp is illuminated, reverts to the " manual" mode and the red trip lamp is illuminated.
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Page 34 . . QUESTION: 048 (1.00) Given the following conditions:
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A loss of decay heat removal has occurre The RCS is NOT exposed to atmosphere (Filled and vented).
-- RCS incore temperature 200 degrees RCS pressure 240 psi Select the method to remove decay hea An OTSG with a Reactor Coolant Pump Feedwater (Emergency or Main) Bypass valves (Turbine or Atmospheric). Spent Fuel Cooling System with a valve line-up to supply the Decay Heat Syste The BWST using a gravity feed metho The BWST using the Reactor Building Spray pump QUESTION: 049 (1.00) Given the following conditions:
--
The "A" Emergency Diesel Generator (EDG) is running and supplying loads on the 4160 ES Bu Panel DPDP-6A has been de-energized, which causes a loss of DC control power to "A" ED Select the resultant operating condition of the "A" ED The "A" EDG: a. trips with no alarm indication b. trips and all attendant alarms operate as require c. remains running with the shut down relay defeate d. remains running with the emergency stop relay defeate i l l l l
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J t QUESTION: 050 (1.00) ,
'Given the following conditions:
<
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Emergency Diesel Generator 1B (EGDG-1B) is running parallel with the Grid and loaded to 2610 K l Choose the method to INCREASE KW loading on EGDG-1B and DECREASE MVARs l OU ; Go to " Raise" on the " GEN SPEED" switc Go to " Lower" on the "EXC VOLT ADJ DIESEL GENERATOR B" switc i Go to " Lower" on the " GEN SPEED" switc ' Go to " Lower" on the "EXC VOLT ADJ DIESEL GENERATOR B" switch, i Go to " Lower" on the " GEN SPEED" switc l Go to " Raise" on the "EXC VOLT ADJ DIESEL GENERATOR B" switc ;
-; Go to " Raise" on the " GEN SPEED" switc .I Go to " Raise" on the "EXC VOLT ADJ DIESEL GENERATOR B" switc :
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. . QUESTION: 051 (1.00) Given the following conditions:
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Refueling operations are in progres A new fuel assembly is being transferred from the spent fuel pool to the fuel transfer cana The running "A" Decay Heat pump must be secured for maintenance prior to placing the "B" Decay Heat train in servic Choose the evolution which should NOT be performed while the decay heat train is of The transferring of New fuel from the Spent Fuel Pool to the Fuel Transfer Canal, Increasing the level of the Spent Fuel Pool by adding de-ionized wate Increasing Decay Heat flow above the minimum continuous DH pump flow of 1400 gp Lowering the Reactor Vessel level during refueling operation j l
' REACTOR OPERATOR a *
b,. Page 37 . . QUESTION: 052 (1.00) Given the following conditions:
--
The RCS has been drained down to "MID LOOP OPERATIONS."
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"A" Decay Heat Train is in operation at 1400 gp Maintenance is to be performed on Bus "ES A 4160."
Select the required mode of operation for the Decay Heat Syste Isolate both Decay Heat trains until EGDG-1A output breaker maintenance is complet Cross-tie the 4160 VAC ES Bus 3A with 4160 VAC ES Bus 3 Place the "B" Decay Heat Train in service and the "A" Decay Heat Train in standb Switch the breaker control handle for DHV-3, Inboard Decay Heat Suction Isolation, to the closed positio l
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QUESTION: 053 (1.00) Given the following conditions:
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Reactor Building pressure is 4 psi The Reactor Coolant Drain Tank (RCDT) rupture disc has been rupture A Pressurizer Code Safety valve is liftin Pressurizer Temperature is 656 degrees What is the tail pipe temperature for the lifting safety? degrees degrees degrees degrees )
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Page 38 . . QUESTION: 054 (1.00) Given the following conditions:
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A normal pressurizer start-up is in progres Pressurizer pressure is 100 psig and being vented to the Reactor Coolant Drain Tank (RCDT).
What are the indications that a steam bubble has been drawn in the Pressurizer and venting is complete? Pressurizer Pressure is decreasing RCDT pressure is increasing RCS pressure is constant RCDT pressure is constant Pressurizer pressure is constant RCDT pressure is increasing RCS pressure is increasing RCDT pressure is decreasing QUESTION: 055 (1.00) Given the following conditions:
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A low level alarm is illuminated for the Nuclear Services Closed Cycle Surge Tank (SWT-1).
-- Nuclear Services Closed Cycle Cooling Pump (SWP-1B) has auto starte An Engineered Safeguards (ES) signal, RCS Pressure Less than 1500 psig, is indicated on all channel Other than NORMAL ES loads, what ADDITIONAL SW heat load is isolated in these circumstances? a. The Reactor Building Fan Assemblie b. The Control Rod Drive Motor Assemblie c. The operating Make-up Pump The Reactor Coolant Pump ; l
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. . QUESTION: 056 (1.00) With an Engineered Safeguard signal present, what system becomes an ADDITIONAL heat load on the Nuclear Services Closed Cooling System? The Reactor Building Fan Assemblie The Control Rod Drive Motor Assemblie The operating Make-up Pump The Reactor Coolant Pump QUESTION: 057 (1.00) Given the following conditions:
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A Reactor / Turbine trip has occurre The Turbine Governor valves indicate " closed" on the turbine EHC control syste "B" OTSG has developed a small steam leak and is controlling at 800 psi "A" OTSG is controlling at 800 psig and does NOT have a steam lea Why are the OTSG's controlling at the same pressure? ICS selected header pressure control is select to "B" OTS The steam leak on "B" OTSG is located inside the Reactor Buildin The "B" OTSG pressure has caused main feedwater to overfeed "A" OTS The OTSGs are cross-connected through the turbine throttle valve .
& ^ * REACTOR OPERATOR w W Page 40 . .
QUESTION: 058 (1.00) Given the following conditions:
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Instrument air pressure has decreased to 83 psig and is constan VP-580, " Plant Safety Verification Procedure" is being performe A " differential flow" alarm on the Control Rod Drive Mechanisms REQUIRED isolation of the Nuclear Services Closed Cycle Ccaling System (NSCCCS).
Select the REQUIRED operator action once the NSCCCS is isolate Trip theVerification."
Status reactor and concurrently perform EOP-02, " Vital System Stop all 4 Reactor Coolant Pump Trip both Main Feedwater Pump Isolate Instrument Air Valve IAV-30 (cross-connect between Instrument Air and Service Air).
QUESTION: 059 (1.00) Given the following conditions:
-- --
Reactor power is 100%. Group 7 rods, the controlling group, begins to move IN continuousl Rod motion stops when group 7 is at their in-limi Proper rod sequencing was observe Reactor power is 65%. What core operating limit is closest to exceeding its limit? Quadrant Power Tilt Regulating Rod Insertion Limits Rod Program i Axial Power Shaping Rod Insertion Limits
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' QUESTION: 060 (1.00)
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Given;the following conditions: '
! --
Group 7, Rod 5 sticks while withdrawing rods for criticality
--
An asymmetric rod fault lamp is illuminated on the Diamond i Pane ;
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Select the position of Group 7, Rod 5 when the fault lamp illuminate Exactly 9' inches misa11gned from another rod in Group . l Exactly 6.0% index difference from another rod in Group ' Greater than 9 inches misaligned from the Group 7 averag i Greater than 9% index difference from another rod in Group l
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i QUESTION: 061 (1.00) ,
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Given the following conditions:
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An Asymmetric Rod Fault alarm is illuminate *
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The Out-Inhibit Lamp is illuminate +
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Group 7 rod 3 rod bottom light is illuminate Choose the " operating condition" of the IC ' A plant runback is in progress to less than 60% powe The Diamond Panel has shifted to " manual" and the Auto-Inhibi , light is illuminate I e The Diamond Panel has shifted to " manual" and the Sequence- , Inhibit light is illuminate .;
! Feedwater-Reactor crosslimits are in affect causing a plant .l runback to less than 60% powe !
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. . QUESTION: 062 (1.00) Given the following conditions:
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Group 7, rod 3 (Rod 7-3) had been previously stuc Rod 7-3 has been freed using procedure OP-50 However, the relative rod position indication (RPI) on the Pasition Indication (PI) panel for Rod 7-3 DOES NOT agree
.ith its absolute position indication (API).
What is required to realign Rod 7-3 RPI to match Rod 7-3 API? Move Group 7 rods to Rod 7-3 API and realign all Group 7 rods with Rod 7-3 AP Drive all Group 7 rods to the in-limit and realign the rods to the Zero positio Withdraw all Group 7 rods to the nearest zone indicating lamp and realign all Group 7 rods to the zone reference indicatio Select Rod 7-3 on the Group and Single Select switch, and use the Reset Pulser to align RPI with AP QUESTION: 063 (1.00) Given the following conditions:
--
A small break LOCA has occurred which CANNOT be isolate RCS pressure 650 psi Subcooling Margin (SCM) 52 degrees Select the required status for the Core Flood Tank The Core Flood Tanks should be: isolate vente allowed to inject until SCM is greater than 70 degrees allowed to inject until LPI injection into the core is verifie a'"" d!I
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.. . QUESTION: 064 (1.00) Select the plant parameters which indicate a LOCA is in progress PRIOR to a reactor trip being initiate RCS Pressure is decreasin RCS Temperature is decreasin Secondary Saturation Temperature is decreasin RCS Pressure is decreasin RCS Temperature is constan Secondary Saturation Temperature is constant, RCS Pressure is decreasin RCS Temperature is increasin Secondary Saturation Temperature is decreasin RCS Pressure is constan RCS Temperature is decreasin Secondary Saturation Temperature is constan yUF9 TION: 065 (1.00) Given the following conditions:
--
The reactor has tripped from 100% powe The Subcooling Margin is O degrees RCPs are of Primary T-hot and Primary T-sat are equa Primary T-cold and Secondary T-sat are equa Select the mode of RCS cooling occurring for the present condition Single phase Natural Circulation Forced convection Boiler-condenser Natural Circulation Natural Conduction
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. . QUESTION: 066 (1.00) Given the following conditions:
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A slow seal degradation is indicated on "A" RC "A" Reactor Coolant Pump (RCP) Third Stage Seal temperature is 175 degrees Choose the required IMMEDIATE operator actio Stop the "A" RC Reduce reactor power to less than 72% and stop the "A" RC Verify the high temperature condition and that RCP cooling water and seal injection is adequat Trip the Reactor, isolate the Reactor Building, and immediately stop the "A" RC QUESTION: 067 (1.00) Given the following conditions:
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Due to a malfunction on one RCP, the plant is in "Three Pump" operation The proper precautions have been taken and the plant has returned to 75% powe Select the OTSG level for the RCS loop which has 2 RCPs operatin OTSC level is equivalent to the 4 pump: % power level, % power leve % power leve % power leve : I I
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QUESTION: 068 (1.00) ;
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Given the following conditions:
. --
A loss of Subcooling Margin has occurred, l ,
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Reactor Coolant System (RCS) pressure is 400 psi }
--
EXCEPT for Reactor Building, Spray / Chemical Addition, all .j Emergency Safeguards Actuation Systems (ESAS) have i initiate i j Choose the ESAS component (s) allowed to be repositioned out of their , ESAS required positio , The RCP seal return and SW cooling valves to the running RCP f
! One HPI pump can be stoppe { EFIC can be bypassed and OTSG levels maintained at 50% in the operating rang ;
'i ' If LPI flow is less than 500 gpm, then stop both LPI pum t i
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QUESTION: 069 (1.00) The Diverse Scram System (DSS) provides a reactor trip signal as a back i
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up for the Reactor Protection System (RPS).
Select the INSTRUMENT and PARAMETER used for the input to the DS RPS Channels A & B reactor coolant temperature j Remote Shutdown Panel reactor coolant pressure l Emergency Safeguards Actuation System reactor building pressure RPS Channels C & D High Powe"/ Flux
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.. . QUESTION: 070 (1.00) Given the following conditions:
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Reactor power is 28%.
--
Main Feedwater has suddenly decreased to 15% feed flo Select the expected automatic plant respons The Reactor is Feedwater Limited (crosslimits) and feedwater flow is increased to 28%. The Feedwater is Reactor Limited (crosslimits) and feedwater flow is increased to 28%. The operating Main Feedwater pump shifts from speed control to
" delta p" contro EFIC initiates and the Main Turbine is trippe QUESTION: 071 (1.00)
The function of the ATWS Mitigation System Actuation Circuitry (AMSAC) is to provide: a diverse means to initiate EFIC and trip the main turbine upon a loss of feedwater during power operatio a reactor trip signal independent of the Reactor Protective System on indication of high RCS pressur a diverse means to balance reactor power and Main Feedwater flow during power operatio a reactor runback signal independent of the Integrated Control System on indication of RCS flow mismatc i a
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t QUESTION: 072 (1.00) Given the following conditions:
--
A large steam leak has occurre OTSG "A" indicates 590 psi OTSG "B" indicates 460 psi Select the expected EFIC feedwater respons EFIC vector logic will send: open commands to both OTSG feedwater valves and will feed both OTSG a close command to "A" OTSG feedwater valves, an open command to "B" OTSG feedwater valves, and will feed only "B" OTS an open command to "A" OTSG feedwater valves, a close command to "B" OTSG feedwater valves, and will feed only "A" OTS close commands to both OTSG feedwater valves and will NOT feed either OTS ~. _ __ .__
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-QUESTION: 073 (1.00) a Given the following conditions: --
An " Excessive Heat Transfer" transient has occurre What criteria is.used to determine a " Stable OTSG?" [ a.- OTSG pressure'is decreasing ,
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OTSG pressure is at Psat for the RCS Tc OTSG level is constant i OTSG pressure is NOT decreasing , OTSG pressure is at Psat for the RCS Tc : OTSG level is constant , OTSG pressure is NOT decreasing . i OTSG pressure is 100 psi below Psat for the RCS Tc " OTSG level is increasing OTSG pressure is decreasing from the OTSG safety lift setpoint OTSG pressure is at Psat for the RCS Tc OTSG level is increasing
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QUESTION: 074 (1.00) Given the following condition A station blackout has occurre Select the equipment responsible for cooling the reactor cor The Make-up pumps used as HPI pump , The Electric Driven Emergency Feedwater pum The Decay Heat pumps used as LP.I pump d The Steam Driven Emergency Feedwater pum T
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QUESTION: 075 (1.00) ; Given the following conditions:
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Reactor refueling is in progres EGDG-1A is out of service for maintenanc A Loss of Off-site Power has occurre EGDG-1B FAILED to star Choose the AP/EOP to be entere EOP-12, " Station Blackout." EOP-11, " Loss of Decay Heat Removal." AP-581, " Loss of NNI-X." EOP-4, " Inadequate Heat Transfer."
QUESTION: 076 (1.00) Given the following conditions:
--
Reactor power was at 100%.
--
A station blackout has occurre A loss of subcooling margin has occurre Select the applicable MAXIMUM RCS cooldown rate, The maximum achievable rate Less than or equal to 100 degrees F per hour Less than or equal to 50 degrees F per hour Less than or equal to 25 degrees F per hour
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. . QUESTION: 077 (1.00) Given the following conditions:
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The NNI-X white indicating light on the Redundant Instrument Panel has extinguishe Annunciator alarm, "NNI-X Power Failure," is illuminate A red trip light on the SASS module indicates that: the channel has lost powe the channel has NNI-Y power being supplied to i the channel has swapped to the alternate instrumen MUV-31, Pressurizer Level Control, has transferred to manua QUESTION: 078 (1.00) The Shift Supervisor has been notified by one of the Fossil Units of a large, chlorine gas releas The Crystal River 3 Control Room has become uninhabitabl The Control Room Operators shall: notify plant personnel and evacuate the Control Roo trip the reactor and evacuate the Control Roo commence a plant shutdown and don air pack notify plant personnel and don air pack ' =-
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. . QUESTION: 079 (1.00) During a " Shutdown From Outside the Control Room," the Chief Nuclear Operator has locally opened CRD Breakers A, B, C, and , The Chief Nuclear Operator must also verify: Source Range Nuclear Instrumentation on scale and DECREASIN CRD groups 1 through 7 are fully inserte boron concentration is INCREASIN a negative 1/3 dpm start-up rate on the Source Range Nuclear Instrumentatio QUESTION: 080 (1.00) Given the following conditions:
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A plant ecoldown has been complete The Reactor Coolant System is at the refueling boron concentratio What 1 irements must be met prior to opening both doors of Reactor Buildin; c ersonnel and equipment hatches? The requirements in: a. the ODCM, "Off-site Dose Calculation Manual."
b. AP-880, " Fire Protection."
c. AI-502, "Defueled Plant Operations."
d. OP-409, " Plant Ventilation Systems."
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. . QUESTION: 081 (1.00) Given the following conditions:
--
Incore temperature indicate greater than 20 degrees superheated condition HPI is in operatio MUP-1B and -1C are on a single BWST suction lin BWST Level is 20 fee Select the suction source water supply for the HPI pump HPI suction is aligned to the: Make-up Tan Borated Water Storage Tank (BWST). LPI system discharg Reactor Building sum QUESTION: 082 (1.00) Given the following conditions:
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Incore temperatures indicate greater than 20 degrees superhea Temperature of the cladding indicates within the " Region IV" are Choose the requirements for Reactor Coolant Pump (RCP) operatio Start ONE RCP per loop when all RCP start permissives are met, Start ALL RCPs when all RCP start permissives are met, Start ONE RCP per loop by bypassing RCP start permissive Start ALL RCPs by bypassing RCP start permissive I
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QUESTION: 083 (1.00)
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Given the following conditions: i
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A reactor trip has occurred from 100% powe i
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The "A" OTSG Main Feedwtater Block valve, FWV-30, has stuck ope l
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The Control Room operators have tripped both Main Feedwater ; pump t Choose the Emergency Feedwater pump (EFW) start signal for the.above l condition i Both EFW pumps start on s " Loss of Main Feedwater pumps." l The steam driven EFW pump, EFP-2, is manually starte ; The electric driven EFW pump, EFP-1 is manually' started,
' Both EFW pumps start on EFIC low level in the OTSG i'
i QUESTION: 084 (1.00) ' Given the following conditions: i
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A reactor trip has occurred from 100% powe OTSG "A" and "B" pressures indicate 1030 psi & Choose how OTSG pressure is being controlle f OTSG pressure is being controlled by: l the Turbine Bypass Valves (TBVs).
i- the Main Steam Safetie : the Turbine Bypass Valves (TBVs) and the Atmospheric Dump [ Valves (ADVs). { the Atmospheric Dump Valves (ADVs) and the Main Steam Safetie i ,
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. .. . QUESTION: 085 (1.00)
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Given the'following conditions:
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A Loss of Off-site Power has occurred coincident with a RCS leak of 20 gpm
--
The reactor trip was assumed to occur due to the Loss of
.Off-site Powe ' --
T-hot temperature is increasing towards saturation temperatur I
--
T-cold temperature has decreased below Tsat in the OTSG '!
!
Select the cause for these indication j The leak is located on the RCS Thot le l The leak is located on the RCS Tcold le A loss of Natural Circulation has occurre , Inadequate Core Cooling has develope i QUESTION: 086 (1.00) Given the following conditions:
--
Reactor Power is 100%.
--
RCS Pressure is decreasin ,
--
Pressurizer Level is 220 inche ;
' --
Pressurizer Temperature is decreasin Make-up Tank level is constan , Choose the cause for these indication [, A Pressurizer steam leak is occurrin . A RCS leak in the letdown system is occurrin > The Pressurizer (PZR) level is below the heater cutoff leve The Pressurizer Spray Valve, RCV-14, is stuck open, i i
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. . QUESTION: 087 (1.00) While performing a reactor start-up, the overlap between the Source and Intermediate Range Nuclear Instruments was 4 decade Choose the cause for this amount of overlap, The Intermediate Range compensating voltage is set to LO The Source Range high voltage in set to HIG The Axial Power Shaping Rods (APSRs) are " Shadowing" the detector A high boron concentration has caused the reactor to be."Over-moderated."
QUESTION: 088 (1.00) Given the following condition:
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An OTSG tube rupture has occurred in ONE OTS The RCS leak rate is approximately 200 gp A Reactor trip has occurre Choose the parameter (s) used to identify the affected OTS The affected OTSG may be identified using: the condenser off-gas monitor, RMA-1 selected main steam header pressur RCS loop Tc temperature OTSG feed flows and OTSG level ]
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. . QUESTION: 089 (1.00) Given the following conditions:
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A reactor trip has occurred from 100% powe "B" OTSG had a stuck open steam safety which has reseate "B" OTSG level indicates 5 inches by EFIC low range indicatio Select the PREFERRED feedwater source and reason for using it to recover the "B" OTS Emergency Feedwater is preferred because recovery with Main Feedwater has not been analyze Emergency Feedwater is preferred because it sprays directly on the OTSG tube Main Feedwater is preferred because it is preheated in the downcomer prior to reaching the OTSG tube Auxiliary Feedwater is preferred because it is preheated in the downcomer prior to reaching the OTSG tube '
* REACTOR OPERATOR Page 57
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QUESTION: 090 (1.00) t Given the following condition:
-- A bulb in the ICS delta Tc station was being replaced when an electrical short occurre Both Main Feedwater Block, Low Load Block, and Start-up valves have shu The reactor has tripped on High RCS pressure.-
Choose the REQUIRED method of feed and level control for the OTSGs with NO operator actio , Main Feedwater is controlling at low level limit Auxiliary Feedwater is controlling at low level limit Emergency Feedwater is controlling at 50% in the operating rang ; Emergency Feedwater is controlling at EFIC low level limit >
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. . QUESTION: 091 (1.00) Given the following conditions:
--
The reactor is in MODE A containment purge is in progres Personnel are in the Reactor Building doing a walkdown of the Reactor Coolant Syste I A HIGti FLOW alarm has been received by RM-A1, the containment purge gas mor.ito Select the REQUIRED Control Room operator actio Isolate the Containment purge supply and exhaust valve Isolate the Reactor Building Vent valves, LRV-70, 71, 72, and 7 Stop the Containment purge supply and exhaust fan Sound the Reactor Building evacuation alarm and shut and lock the Reactor Building Personnel Hatc _ _ __ __
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' REACTOR OPERATOR Pa9e 59 . . QUESTION: 092 (1.00) Given the following conditions:
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Instrument Air pressure indicates 80 psi The HPs have notified the Control Room that NO individuals are on breathing ai NSCCCS valve, SWV-110, " Return from the CRDMs," has failed close The Reactor Building sump level is slowly increasin Select the REQUIRED control room operator actio Trip the reactor and concurrentli perform EOP-02, " Vital Systems Status Verification." Perform a controlled shutdown at the maximum safe rat Trip the reactor coolant pumps and perform EOP-09, " Natural Circulation Cooldown." Sample the Reactor Building sump to identify type of leakag QUESTION: 093 (1.00) Given the following conditions:
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Reactor Building sump level is increasin Pressurizer level is constant at 220 inche Make-up Tank level is decreasin Select the cause for these indication A small Main Steam leak inside the Reactor Buildin A small RCS leak on the letdown system in the Auxiliary Buildin A RCS leak within the capacity of normal make-u l l A leaking Pressurizer Code Safet l
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. . QUESTION: 094 (1.00) Given the following conditions:
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A Loss of Off-site Power has occurre The running Make-up pump is NOT the ES selected pum Natural Circulation has been established and the plant is stable in Mode What is REQUIRED to re-establish Make-up and Purification and what precaution (s) should be taken with the REQUIRED action, The Make-up pump must be restarted and RCP seal injection must be closed and slowly re-establishe Re-establish letdown by bypassing the High temperature interlock and slowly increase flow back to the desired amoun Re-establish RCP seal controlled bleed-off and ensure the proper RCP seal staging is observe Place the Pressurizer Control Valve, MU-31, in automatic and ensure the Pressurizer Spray Valve, RCV-14, is shu QUESTION: 095 (1.00) A large tear has developed in the expansion joint between the turbine and condenser. The Reactor has tripped due to a turbine tri The turbine trip was on "a loss of Main Condenser vacuum".
Identify the method for controlling OTSG pressure once hot shutdown conditions are met (RCS Tave = 532 degrees F), Turbine bypass contro Periodically lifting the OTSG steam safetie Feeding the OTSGs with emergency feedwate Atmospheric dump valve contro .. . - . - . - .- . . . . o v O Page 61
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QUESTION: 096 (1.00) ,
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Given the following Conditions:
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Reactor power is 100% with all ICS control stations in .[
" automatic." 1 --
SASS is NOT operationa ;
--
SELECTED turbine header pressure fails HIG l Select the expected feedwater response for the above condition ' Feedwater flow will: i cycle (increase and decrease constantly). decreas I
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QUESTION: 097 (1.00)- I
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RCS pressure has decreased to 1450 psi Reactor building pressure is 1 psi ESAS has actuate ;
--
Make-Up Tank outlet check valve,.MUV-65 has seate ... Identify the Make-Up Tank- (MUT-1) level response and cause for the l chang l f Make-Up Tank level: i increases because of RCP Control Bleed Off Return flo ! decreases because of HPI Injection flow into the Reactor Cor t increases because of Make-Up Pump Recirc Return flo j decreases because Letdown flow is isolate .[
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' QUESTION: 098 (1.00)
..
Which of the following conditions will cause an Out-Inhibit lamp to be illuminated? A 1.5'dpm start-up rate in the source rang ' A 3.0 dpm start-up rate in the intermediate rang Reactor power is 50% power and safety. rods are NOT at.the "out limit." The Reactor is greater than 60% power and a 7-inch adymmetric fault exist : ,
'
t QUESTION: 099 (1.00) , t Relative Rod Position differs significantly from Absolute Rod Position indicatio . Identify the cause for the difference between the two indications, A Reactor Trip has occurre A loss of ICS auto power has occurre Control Rod "out" motion with an "in" comman r A Sequence-Inhibit exist :;
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. . QUESTION: 100 (1.00)
,
The Reactor has tripped from 100% power and natural circulation has been establishe Which of the following is an indication that natural circulation has been established? RCS delta T is 90 degrees F and increasin Cold leg temperatures approach saturation temperature for pressurizer pressur RCS flow is indicated by NNI indicatio Incore thermocouple temperatures stabilize, and are tracking T-ho (********** END OF EXAMINATION **********)
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. . ANSWER: 001 (1.00) REFERENCE: Conduct of Operations, AI-500, Section 4.3.1.2, page 4 Lesson Plan, ROT-5-3 [4.1/3.9] 194001A102 ..(KA's) ANSWER: 002 (1.00) . REFERENCE: 10CFR20.101 ROT-5-43, LO B [2.8/3.4)
. . . I 194001K103 ..(KA's) l
. . ANSWER: 003 (1.00) ;
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' REACTOR OPERATOR a '"
O Page 65 ' REF'ERENCE: AI-412, " Verbal Communication Guideline."
LP, ROT-4-92, LO G [3.0/3.2] 194001A104 ..(KA's) ANSWER: 004 (1.00) REFERENCE: AI-412, " Verbal Communication Guidelines," page 6, step 4. ROT-5-3 [3.6/3.8] 194001A105 ..(KA's) ANSWER: 005 (1.00) REFERENCE: AI-500, " Conduct of Operations," page 48, step 4.4. LP, ROT-5-38 (3.4/3.4] 194001A106 ..(KA's)
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;
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' ANSWER: 006- (1.00)
. t
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EM-103, " Operation and Staffing of the CR-3 Control Room During Emergency Classifications," page 2, Section 3. LP, ROT-5-34 ;
'
[3.1/4.4]
,
194001A116 ..(KA's)
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ANSWER: 007 (1.00) .- REFERENCE: EM-202, " Duties of the Emergency Coordinator," page 53, enclosure LP, ROT-5-34
[3.1/4.4]
194001A116 ..(KA's)
-ANSWER: 008 (1.00)
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REFERENCE: AI-1600, "Alara Program Manual," page 9, Section 4. LP, ROT-5-38 i
(3.3/3.5]
194001K104 ..(KA's) ANSWER: 009 (1.00) REFERENCE: CP-115, " Nuclear Plant Tags and Tagging Orders," page 13, Step 4. LP, ROT-5-40
[3.7/4.1]
194001K102 ..(KA's) ANSWER: 010 (1.00) REFERENCE:
'
CP-115, Nuclear Plant Tags and Tagging Orders," page 23, step 4.11. LP, ROT-5-40
[3.7/4.1]
194001K102 ..(KA's)
) ' REACTOR OPERATOR Page 68
- ,
ANSWER: 011 (1.00) REFERENCE: AI-500, " Conduct of Operations," page 10, step 4.1. LP, ROT-5-38
[3.6/3.7)
194001K101 ..(KA's) ANSWER: 012 (1.00) REFERENCE: CP-115, " Nuclear Plant Tags and Tagging Orders," page 12, Step 4. LP, ROT-5-40
[3.6/3.7)
194001K107 ..(KA's) ANSWER: 013 (1.00) l
,
' O :O Page 69- ' REACTOR-OPERATOR ); REFERENCE: .OP-503, " Plant Computer System," page 6, step 4.1.1.
p-l-LP,-ROT-4-21'
[3.1/3.4] -194001A115 ..(KA's)
ANSWER: 014 _(1.00)
' l *'
REFERENCE: ! l OP-210, step 3. 'LP ROT-5-02, page 24
. -
[4.2/4.3] f 001000K518 ..(KA's) {
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REFERENCE: ROT-4-28, Control Rod drive (electrical), Fig.1re 3 : ROT-4-28, LO i
[4.5/4.4] *
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. . ANSWER: 016 (1.00) REFERENCE: OP-302, "RC Pump Operation," Enclosure 1 and Check with facilit [3.5/3.9]
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[3.6/3.7]
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[3.8/3.9]
004000A401 ..(KA's)
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' REACTOR OPERATOR fh Page 71
. . ANSWER: 019 (1.00) W ot, tK ppn-
'N{PfU REFERENCE:
ROT-4-14, " Engineered Safeguards Actuation System," Tables I & I LO B [4.2/4.4] 013000K106 ..(KA's) ANSWER: 020 (1.00) REFERENCE: ROT-4-13, " Engineered Safeguards Actuation System," page 13, 1.4.4 & 1. LO ROT-4-13, C3
[3.9/3.8)
013000G009 ..(KA's) ANSWER: 021 (1.00) i
* REACTOR OPERATOR Page 72
REFERENCE: OP-202, " Plant Heatup," page 59, step 4.5.13 ROT-5-02, page 23, LO [3.9/4.0] 013000G013 ..(KA's) ANSWER: 022 (1.00) REFERENCE: ROT-4-12, " Reactor Protection System," page 1 LO B [4.1/4.2] 015000K101 ..(KA's) ANSWER: 023 (1.00) REFERENCE: ROT-4-28, "CRD (Electrical) ," page 4 LO B1 [3.7/3.9) 015000K402 ..(KA's)
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~* REACTOR OPERATOR .O -
Paga 73-u- . ANSWER: 024- (1.00) REFERENCE: ROT-3-03, " Natural Circulation," page 5, stepR LO B4
[3.4/3.7]
01702GK401 ..(KA's) ,
,
(1.00)
'
' ANSWER: 025 ,
REFERENCE: . t
: ROT-4-13, " Engineered Safeguards I.ctuation System,".page 11, Step'1; l LO C6 [3.3/3.5]. j 022000G007 ..(KA's) f i
ANSWER: 026 (1.00) : !
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REFERENCE: ROT-4-14, " Integrated Control System," page 32, step 3. LO B [3.3/3.5] 059000K402 ..(KA's) ANSWER: 027 (1.00) REFERENCE: ROT-4-14, " Integrated Control System," page 36, step 3.4.12 LO B [3.2/3.2] 059000K107 ..(KA's) ANSWER: 028 (1.00) REFERENCE: ROT-4-15, " Emergency Feedwater and EFIC," page 19 LO B13
[3.9/3.9]
061000A303 ..(KA's) i
* REACTOR OPERATOR Page 75
. . ANSWER: 029 (1.00) REFERENCE: ROT-4-15, " Emergency Feedwater and EFIC," page 5 B3
[3.7/3.7]
061000K202 ..(KA's) ANSWER: 030 (1.00) REFERENCE: Emergency Operating Procedure, "EOP Entry Conditions," page 1 ROT-5-96, page 2, LO B [4.4/4.6] 061000K301 ..(KA's) ANSWER: 031 (1.00) >
.-. '# ' '" * REACTOR OPERATOR Page 76
REFEFENCE: ROT-4-59, Liquid Waste," page 50, step 2.2.37 LO F. [3.6/3.6] 068000A302 ..(KA's) ANSWER: 032 (1.00) REFERENCE: Tech Specs, 3.9.9, page 3/4 9-9, " Containment Purge."
LO ROT-5-1, B10
[3.6/3.8]
071000A302 ..(KA's) ANSWER: 033 (1.00) REFERENCE: Technical Specifications, 3.4. LO ROT-4-25, " Radiation Monitoring System," page 28, section [3.1/3.3] 072000G004 ..(KA's)
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ANSWER: 034 (1.00) , REFERENCE: Ch'eck With Facilit ,
[3.4/3.6]
072000A101 ..(KA's)
.
:
ANSWER: 035 (1.00) . i REFERENCE: ROT-3-03,." Natural Circulation," page 6, Section 4. i
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[4.3/4.5]
002000A402 ..(KA's) ANSWER: 036 (1.00) ,
. *
REFERENCE: i ROT-5-02, " Plant Operating Procedures." { LO Terminal RO Objective : i'
[3.6/3.9]
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. . ANSWER: 037 (1.00) REFERENC1': EOP-03, page 4, Rules 1 & ROT-5-96, B [3.9/4.0] 006000G013 ..(KA's) ANSWER: 038 (1.00) REFERENCE: ROT-4-13, " Engineered Safeguards Actuation System," page 31 LO B [4.1/4.3] 006000K413 ..(KA's) ANJWER: 039 (1.00) r
' REACTOR OPERATOR a "
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'
RENERENCE: ROT-4-13, " Engineered Safeguards Actuation System," page 9, Section 1. LO B [3.6/3.7] 006020A303 ..(KA's) ANSWER: 040 (1.00) REFERENCE: ROT-4-60, " Reactor Coolant System," page 23 LO B4
[3.6/3.5]
010000A302 ..(KA's) ANSWER: 041 (1.00) REFERENCE: ROT-4-60, " Reactor Coolant System," page 2A LO B6
{4.0/3.8)
010000A403 ..(KA's) _ -.- j
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* REACTOR OPERATOR
. . ANSWER: 042 (1.00) REFERENCE: ROT-4-60, " Reactor Coolant System," page 2 LO 88
[3.3/3.7]
011000K401 ..(KA's) ANSWER: 043 (1.00) REFERENCE: ROT-4-60, " Reactor Coolant System," Check with facilit [3.1/3.1) 011000K604 ..(KA's) ANSWER: 044 (1.00) . . "# f x> (.,ll ;
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REACTOR OPERATOR Page.81-REFERENCE: ROT-4-12, " Reactor Protection. System," page 2 !
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LO B [3.1/3.1] 012000K603 -..(KA's) ANSWER: 045 (1.00)
: ,
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' .[3.1/3.5]
012000K603 ..(KA's)
:
ANSWER: 046 (1.00) [
, REFERENCE: l ROT-4-9, "Non-Nuclear Instrumentation," page 40 i LO B i
[3.4/3.4] i
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[3.2/3.3]
016000 GOO 4 ..(KA's) ANSWER: 048 (1.00) REFERENCE: EOP-11, " Loss of Decay Heat Removal," page 7, step LO ROT-4-32, "OTSG," B6 .
[3.4/3.6)
035000G001 ..(KA's)
-ANSWER: 049 (1.00) +
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Page 83
+
a 4-~ , REFERENCE: .
'
OP-700E, "125/250 VDC Distribution Panels," page 4 :LO Check with facilit ;
[3.7/4.1)
'
063000K301 ..(KA's) F 1 ANSWER: 050 (1.00) I l t REFERENCE: OP-707, " Operation of the ES Emergency P_esel Generators," page 38 LO B4
[3.7/3.7)
064000G009 ..(KA's) - ANSWER: 051 (1.00)
:
' i REFERENCE: !
STS 3/4. ' LO ROT-4-54, " Decay Heat Removal," B11
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. . REFERENCE: ROT-4-54, " Decay Heat Removal," page 1 LO B13
[3.5/3.6)
005000G001 ..(KA's) ANSWER: 053 (1.00) REFERENCE: ROT-4-60, " Reactor Coolant System."
LO B3/B6
[3.9/4.2]
007000A201 ..(KA's) ANSWER: 054 (1.00) REFERENCE: ROT-4-60, " Reactor Coolant System." i LO B4
[3.1/3.4)
007000K502 ..(KA's)
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. . ANSWER: 055 (1.00) REFERENCE: ROT-4-56, " Nuclear Services Closed Cycle Cooling System," page 1 LO B7
[3.2/3.5)
008000A202 ..(KA's) ANSWER: 056 (1.00) REFERENCE: ROT-4-56, " Nuclear Services Closed Cycle Cooling System," page 1 LO B7
[3.6/3.7)
008030A304 ..(KA's) ANSWER: 057 (1.00) r
,. ~ 'REACIOR OPERATOR Page 86 . .
REFERENCE: ROT-4-66, " Main, Reheat Steam and MSR's," page 24 LO G2
[3.4/3.6]
045000A304 ..(KA's) ANSWER: 058 (1.00) REFERENCE: ROT-5-84, " Loss of Instrument Air."
LO Check with facilit [3.1/3.1] 078000G015 ..(KA's) ANSWER: 059 (1.00) REFERENCE: ROT-5-67, "AP-525 Continuous Control Rod Motion."
[3.2/3.6] 000001K122 ..(KA's) ANSWER: 060 (1.00) * O*
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[3.6/3.3]
000003A103 ..(KA's) ANSWER: 061 (1.00) REFERENCE: ROT-4-28, " Control Rod Drive, Electrical," pages 35 and 3 LO B17
[3.6/3.6)
000003G009 ..(KA's) ANSWER: 062 (1.00) REFERENCE: ,
ROT-4-28, " Control Rod Drive, Electrical," page 10 LO B14, 15, and 1 [3.4/3.4] 000005A105 ..(KA's) l
. (~) (,~~) ~ 'REACIOR. OPERATOR Page 88-
. . t ANSWER: 063 (1.00)
' ,
'
REFERENCE: ROT-3-21, " Loss of Coolant Accidents," page 3 LO B [4.2/4.2] 000011A115 ..(KA's) ANSWER: 064 (1.00) . REFERENCE: ROT-3-20, " Symptom Oriented Procedure Philosophy," page 13 LO B (3.7/3.7] 000011A213 ..(KA's) ANSWER: 065 (1.00)
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. . REFERENCE: ROT-3-03, " Mitigating Core Damage," page 8 LO B5
[4.1/4.4]
000011K101 ..(KA's) ANSWER: 066 (1.00) REFERENCE: OP-302, "RC Pump Operation," page 2 LO B9
[3.0/3.5]
000015A201 ..(KA's) ANSWER: 067 (1.00) REFERENCE: ROT-4-14, " Integrated Control System," page 6 LO B [3.1/3.2] 000015G007 ..(KA's)
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Page 90 ' _ j
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! ANSWER: 068- (1.00). REFERENCE:
EOP-03, " Inadequate Subcooling Margin," page 7, step LO' ROT-5-85, page 8
[4.0/4.2) '{
_
:
000026K303 ..(KA's)
.
.
ANSWER: 069 (1.00) ;
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[4.2/4.5]
000029K301 ..(KA's) 1
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. . . -. REFERENCE: ROT-4-12, " Reactor Protection System," page 4 l 14 C [4.4/4.5] 000029A209 ..(KA's) : F
~ ANSWER: 071 (1.00) ; REFERENCE:
ROT-4-12, " Reactor Protective System," page 43 LO C ,
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[3.8/4.0)
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000029G007 ..(KA's)
.
ANSWER: 072 (1.00)
. !
REFERENCE: - t ROT-4-15, " Emergency Feedwater and EFIC," page 27 , LO B15 ' f
[4.6/4.6] .
000040A101 ..(KA's) -j a o e f
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EOP-05, " Excessive Heat Transfer," page 25 .I LO ROT-5-94 t
[4.0/4.1]
000040A106 ..(KA's) , ANSWER: 074 (1.00) , REFERENCE': ROT-5-100, "EOP Tab 12, Station Blackout," page 6 LO B3
[4.1/4.4]
000055K102 ..(KA's)
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. . REFERENCE: EOP-01, " Entry Conditions."
LO ROT-5-96, "EOP TABS ... " B (4.1/4.1] 000055G011 ..(KA's) ANSWER: 076 (1.00) REFERENCE: EOP-12, " Station Blackout," page 9 LO ROT-5-100, "EOP-12...," page 10, B3
[4.4/4.6]
000055A202 ..(KA's) ANSWER: 077 (1.00) REFERENCE: ROT-5-81, " Loss of NNI-X," page 5 & LO CWF
[3.2/3.6]
000057A214 ..(KA's)
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REACTOR. OPERATOR O' O Page 94
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' ANSWER: 078 (1.00)
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REFERENCE:
,
7AP-513, " Toxic Gas," page 7.
' ROT-5-66, " Toxic Gas Actuation," B .;
[3.3/4.1] ;
.
i 000068G001 ..(KA's)
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ANSWER: 079 (1.00)
,
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[3.8/4.2] q 000068K301 ..(KA's) ;
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Pag 3 95 , . REFERENCE: ROT-4-63, " Containment Systems and Ventilation," page 39 LO B5
[3.5/3.6)
000069G007 ..(KA's) ANSWER: 081 (1.00) REFERENCE: EOP-07, " Inadequate Core Cooling," page LO ROT-3-25, B3
[3.9/4.1]
000074K204 '
..(KA's)
ANSWER: 082 (1.00) REFERENCE: EOP-07, " Inadequate Core Cooling," page 1 LO ROT-3-25, B3
[3.6/3.9]
000074A106 ..(KA's) ANSWER: 083 (1.00) t' ) -
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. . REFERENCE: ROT-4-15, " Emergency Feedwater and EFIC," page 1 LO B8
[4.0/4.6]
000007K301 ..(KA's) ANSWER: 084 (1.00) REFERENCE: EOP-02, " Vital System Status Verification," page 13 LO ROT-5-96, "EOP Tabs...," B3
[3.7/3.7]
000007A110 ..(KA's) ANSWER: 085 (1.00) REFERENCE: ROT-3-03, " Natural Circulation," page LO B4
[4.2/4.7]
000009K101 ..(KA's)
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,
000027A101 ..(KA's) i
' ANSWER: 087 (1.00) :
1 . REFERENCE: l i ROT-4-10, " Nuclear Instrumentation," page 13 & 1 . LO B8 !
(3.1/3.4]
000033A211 ..(KA's) ,
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. ROT'-3-24, " Steam Generator Tube Rupture," page 'LO B4
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[4.4/4.6] [
:l 000038A203 7
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l i ANSWER: 089 (1.00) ,
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ROT-5-96, "EOP Tabs ..., " page 68 ' LO B5 r i
[3.45/4.2] ,
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ANSWER: 090 (1.00) I i REFERENCE: ROT-5-96, "EOP Tabs ...," page 12 ; LO B5 '
[4.4/4.6]
:
000054K304 ..(KA's)
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[3.3/3.8)
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000060G005 ..(KA's)
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ANSWER: 092 (1.00) ~i i j t i
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[3.7/3.9) j
'000065K308 ..(KA's)
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000028A210 ..(KA's) ' ANSWER: 094 (1.00)
> ,
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- [3.8/3.9]
000056A203 ..(KA's) - ANSWER: 095 (1.00) REFERENCE: t ROT-4-66, " Main and Reheat Steam and MSR's," page 4 > LO G3 g
[3.4/3.5]
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REFERENCE: .;
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[3.4/3.7) ,
,
039000A204 ..(KA's)
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ROT-4-52, "Make-Up and Purification," page 39 3 LO B19-C ,
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ANSWER: 098 (1.00)
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:
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'* -RENCTOR OPERATO Page102-
.. REFERENCE: ROT-4-28, "CRD Electrical," page 36 LO B17
[3.7/3.8]
001000K407 ..(KA's) ANSWER: 099 (1.00) REFERENCE: ROT-4-28, "CRD Electrical," page 17 LO B14\16
[3.9/3.8]
014000A205 ..(KA's) ANSWER: 100 (1.00) REFERENCE: ROT-3-03, " Natural Circulation," page 5 LO B4
[3.6/3.8]
017020A301 ..(KA's)
(********** END OF EXAMINATION **********)
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.. l f s UREACTOR OPERATOR ~' ..Page 1-
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.,- 4 .- ANSWER KEY-l l- .
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. MULTIPLE CHOICE 023 '001 'a 024 c '002 b 025 d 003 c 026 d 004 b' 027 a 1005 b P
006 d 029 c t
'007 c 030 a 008 c .f
.
009 a 032 a
.010 b 033 b .f ,
011 c 034 d ! 012 a 035 d ' 013 c ! i 014 d 037 b ;
.
015 b 038 a ,
-016 b 039 c 017 c 04 d l 018 c ' -
j A h, .B 041 b -
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019 042 d KA (( l
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020 a 043 a 021 b 044 b ; 022 a 045 b ' i
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Page 2
- * ANSWER KEY 046 a 068 a MULTIPLE CHOICE 069 b 047 c 070 d 048 a 071 a 049 c 072 c 050 a 073 b 051 b 074 d 052 c 075 b 053 c 076 a 054 b 077 c 055 d 078 d 056 a 079 b 057 d 080 a 058 a 081 c 059 b 082 d 060 c 083 a 061 a 084 c 062 d 085 c 063 a 086 d 064 b 087 a 065 c 088 d
066 c 089 a 067 d 090 d
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Page 3
. . ANSWER KEY 091 c MULTIPLE CHOICE 092 a 093 c 094 a 095 d 096 b 097 a 098 b 099 a 100 d (********** END OF EXAMINATION **********)
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TEST CROSS REFERENCE a'" Page 1
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t RO Exam PWR React or Organized by Question Number QUESTION VALUE REFERENCE 001 1.00 9000001 002 1.00 9000002 003 1.00 9000003 004 1.00 9000005 005 1.00 9000006 006 1.00 9000007 007 1.00 9000008 008 1.00 9000009 009 1.00 9000010 010 1.00 9000011 011 1.00 9000013 012 1.00 9000014 013 1.00 9000015 014 1.00 9000021 015 1.00 9000022 016 1.00 9000024 017 1.00 9000025 018 1.00 9000026 019 1.00 9000027 020 1.00 9000028 021 1.00 9000029 022 1.00 9000031 023 1.00 9000032 024 1.00 9000033 025 1.00 9000034 026 1.00 9000035 027 1.00 9000036 028 1.00 9000037 029 1.00 9000038 030 1.00 9000039 031 1.00 9000040 032 1.00 9000041 033 1.00 9000042 034 1.00 9000043 035 1.00 9000044 036 1.00 9000045 037 1.00 9000046 038 1.00 9000047 039 1.00 9000048 040 1.00 9000049 041 1.00 9000050 042 1.00 9000051 043 1.00 9000052 044 1.00 9000053 045 1.00 9000054 046 1.00 9000057 047 1.00 9000058 048 1.00 9000060 049 1.00 9000062 L
TEST CROSS REFERENCE Page 2- .. - RO 10x a m PWR Reactor Organized by Q u e s t i'o n Number
+
QUESTIO VALUE -REFERENCE , 050 1.00 9000063 051 1.00 9000064 -
-
052 1.00 9000065 , 053 1.00 9000066 .
,
054 1.00 9000067 055 1.00 9000068 f 056 1.00 9000069 057 1.00 9000070 058 1.00 9000071 i 059 1.00 9000072 060 1.00 9000073 'i 061 1.00 900007, 062 1.00 9000075 '; 063 1.00 9000076 ;
'
064 1.00 9000077 065 1.00 9000078 066 1.00 9000079 j 067 1.00 9000080 ' 068 1.00 9000081 ' 069 1.00 9000082 070- 1.00 9000083 l 071 1.00 9000084 072 1.00 9000085 073 1.00 9000086-074 1.00 9000087 ! 075 1.00 9000088 : 076 1.00 9000089 077 1.00 9000090 ! 078 1.00 9000091 079 1.00 9000092 080 1.00 9000093 081 1.00 9000094 082 1.00 9000095 083 1.00' 9000096 ! 084 1.00 9000097 085 1.00' 9000100 I 086 1.00 9000101 , 087 1.00 9000103 088 1.00 9000106 089 1.00 .9000107 090 1.00 9000108 '! 091 1.00 9000109 '
;
092 1.00 9000111 093 1.00 9000112 , 094 1.00 9000114 095 1.00 9000120 l
:
096 1.00 9000121 I 097 1.00 9000122 098 1.00 9000123 : I
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TEST CROSS REFERENCE Page 3 . . RO Exam PWR Reactor Organized by Question Number QUESTION VALUE REFERENCE 099 1.00 9000124 100 1.00 9000125 ______ 100.00 ______ ______ 100.00
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TEST CROSS REFERENCE Page 4
. a; RO Exam PWR Reactor Organized by KA Group PLANT WIDE GENERICS ,
QUESTION VALUE KA 001 1.00 194001A102 003 1.00 194001A104 004 1.00 194001A105
%
005 1.00 194001A106 013 1.00 194001A115 006 1.00 194001A116 007 1.00 194001A116 011 1.00 194001K101 010 1.00 194001K102 009 1.00 194001K102 002 1.00 194001K103 008 1.00 194001K104 012 1.00 194001K107 ______ PWG Total 13.00 PLANT SYSTEMS Group I QUESTION VALUE KA 015 1.00 001000K105 098 1,00 001000K407 014 1.00 001000K518 016 1.00 003000A201 018 1.00 004000A401 017 1.00 004000K520 020 1.00 013000G009 021 1.00 013000G013 019 1.00 013000K106 022 1.00 015000K101 023 1.00 015000K402 100 1.00 017020A301 024 1.00 017020K401 025 1.00 022000G007 027 1.00 059000K107 026 1.00 059000K402 028 1.00 061000A303 029 1.00 061000K202 030 1.00 061000K301 031 1.00 068000A302 032 1.00 071000A302 034 1.00 07: 010A101 033 1.00 072000G004 ______ I _ _ . _ _ - _ .
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TEST CROSS REFERENCE Page 5 . ..- RO Exam PWR Reactor Organized by KA Group PLANT SYSTEMS
,
Group I QUESTION VALUE KA PS-I Total 23.00 Group II QUESTION VALUE KA 035 1.00 002000A402 036 1.00 002020K504 037 1.00 006000G013 038 1.00 006000K413 039 1.00 006020A303 097 1.00 006030K402 040 1.00 010000A302 041 1.00 010000A403 042 1.00 011000K401 043 1.00 011000K604 044 1.00 012000K603 045 1.00 012000K603 099 1.00 014000A205 047 1.00 016000G004 046 1.00 016000K101 048 1.00 035000G001 095 1.00 035000G013 096 1.00 039000A204 049 1.00 063000K301 050 1.00 064000G009 ______ PS-II Total 20.00 Group III QUESTION VALUE KA 052 1.00 005000G001 051 1.00 005000K307 053 1.00 007000A201 054 1.00 007000K502 055 1.00 008000A202 056 1.00 008030A304 057 1.00 045000A304 058 1.00 078000G015 ______ PS-III Total 8.00 ______ ______ J
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TEST CROSS REFERENCE Pag 3 6 . + RO Exam PWR Reactor Organi zed by KA Group PLANT SYSTEMS QUESTION VALUE KA PS Total 51.00 EMERGENCY PLANT EVOLUTIONS Group I QUESTION VALUE KA 062 1.00 000005A105 066 1.00 000015A201 067 1.00 000015G007 068 1.00 000026K303 086 1.00 000027A101 072 1.00 000040A101 073 1.00 000040A106 076 1.00 000055A202 075 1.00 000055G011 074 1.00 000055K102 077 1.00 000057A214 078 1.00 000068G001 079 1.00 000068K301 080 1.00 000069G007 082 1.00 000074A106 081 1.00 000074K204 ______ EPE-I Total 16.00 Group II OUESTION VALUE KA 059 1.00 000001K122 060 1.00 000003A103 061 1.00 000003G009 084 1.00 000007A110 083 1.00 000007K301 085 1.00 000009K101 063 1.00 000011A115 064 1.00 000011A213 065 1.00 000011K101 070 1.00 000029A209 071 1.00 000029G007 069 1.00 000029K301 087 1.00 000033A211 088 1.00 000038A203 089 1.00 000054K102
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Page 7
' . -
RO Exam PWR Reactor Organized by KA Group EMERGENCY PLANT EVOLUTIONS Group II QUESTION VALUE KA 090 1.00 000054K304 091 1.00 000060G005 ______ EPE-II Total 17.00 Group III QUESTION VALUE KA 093 1.00 000028A210 094 1.00 000056A203 092 1.00 000065K308 ______
EPE-III Total 3.00 ( ______ ______ EPE Total 36.00 ______ ______ ______ ! Test Total 100.00 N- }}