IR 05000254/1998303

Exam Repts 50-254/98-303OL & 50-265/98-303OL on 980316-20. Exam Results:All Three Applicants Passed All Portions of Respective Exams

ML20216F577
Person / Time
Site:	Quad Cities
Issue date:	04/13/1998
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20216F557	List: IR 05000254/1998303 ML20216F553
References
50-254-98-303OL, 50-265-98-303OL, NUDOCS 9804170123
Download: ML20216F577 (110)

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U. S. NUCLEAR REGULATORY COMMISSION REGION lil Docket Nos: 50-254;50-265 License Nos: DPR-29; DPR-30 Report Nos: 50-254/98303(OL); 50-265/98303(OL)

Licensee: Commonwealth Edison Company (Comed)

Facility: Quad Cities Nuclear Power Station, Units 1 and 2 Location: 22710 206th Avenue North Cordova,IL 61242 Dates: March 16,1998 Examiners: D. R. McNeil, Chief Examiner '

l J. D. Ellis, Examiner -

T. R. Jones, Examiner (In training)

i-l Approved by: Melvyn N. Leach, Chief Operator Licensing Branch

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9804170123 980413 PDR i

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EXECUTIVE SUMMARY Quad Cities Nuclear Power Station NRC Examination Reports 50-254/98303; 50-265/98303 An NRC developed initial operator licensing examination was administered to three license applicants [two Senior Reactor Operator (SRO) applicants, and one Senior Reactor Operator Limited to Fuel Handling (LSRO) applicant].

Results:

All three applicants passed all portions of their respective examinations. The LSRO was issued a license. The two SRO applicants were not issued operating licenses as they have not completed all requirements to be issued a license. Licenses will be issued upon completion of all Quad Cities Nuclear Power Station training program requirements; le, one additional month of responsible power plant experience at Quad Cities Nuclear Power Station. (Section 05.1)

Examination Summarv:

The training and operations departments staff provided significant assistance in preparing the examination for administration to the applicants. (Sections 05.1, O5.2, 05.3 and O5.4)

The applicants appeared well prepared to take the examination. (Sections 05.2, O and 05.4)

Administration of the examination revealed one applicant performance issue needing additional review by the Quad Cities Nuclear Power Station Training and Operations Departments. On two occasions during simulator transients, license applicants reached setpoints, imposed by an administrative procedure, that required a specified action such as a reactor scram or emergency depressurization. The applicants executed those actions even though they had been informed that the action was unnecessary and the plant was being adequately controlled by crew members. Although the simulated reactor was placed in a safe condition in each of the two cases, the transient initiated on the plant was unnecessary and could have been avoided. (Section 05.5.b)

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Reoort Details 1. Operations 05 Operator Training and Qualification 05.1 General Comments Operator initial license examinations were administered at the Quad Cities Nuclear Power Station (QCNPS) to three applicants during the week of March 16,1998. All three applicants passed all portions of their examinations. One applicant was issued a Senior Reactor Operator Limited to Fuel Handling (LSRO) license. The remaining two applicants were not issued operating licenses because they had not completed all requirements to obtain the licens When 10 CFR Part 55 was revised in 1987, the NRC also revised and published Revision 2 of Regu:atory Guide (RG) 1.8 Qualification and Training of Personnel for Nuclear Powerplants, that provided an acceptable method of implementing the revisions to the regulation. For the SRO-licensed position, the RG endorsed the requirements contained in ANSI /ANS-3.1-1981, Selection, Qualification and Training of Personnel for Nuclear Power Plants, subject to the clarifications, additions, and exceptions ir, paragraphs C.1.a through C.1.1 of the RG. Specifically, Section C. states that an applicant for an SRO license should have four years of responsible power plant experience and that at least six months of the nuclear power plant experience should be at the plant for which an applicant seeks a licens The supplementary information published in the Federa/ Registerwith the 1987 rule change required facility licensees that had made commitments that were less than those required by the new rules to conform to the new rules automatically. Apparently the QCNPS training department was unaware that they were required to revise their training l program requirements and submitted license applications for two employees that had the equivalent of only five months of responsible nuclear plant experience at QCNPS. A waiver was granted to allow the applicants to sit for the examination prior to completing ;

this requirement. Upon certification by the licensing official of completion of the one month of QCNPS experience, the licenses will be issue The examination was developed by NRC Region Ill examiners as part of the certification of an individual as an NRC operator licensing examiner The written examination was composed of 100 closed reference questions; each worth one point. The operating test consisted of four administrative Job Performance Measures (JPMs) and two questions, ,

ten operating JPMs, each with two follow-up questions, and two dynamic simulator scenarios. The operating test was open reference except for some JPM follow-up questions. The written examination and operating test were written in accordance with NUREG 1021, Operator Licensing Examination Standards for Power Reactors, Interim Rev 8 January 199 .

The LSRO examination was originally developed by a member of the OCNPS training staff and submitted to Region lli examiners. The examination consisted of a 50 question written examination, five administrative JPMs, five system JPMs and two scenarios. The submitted examination was of high quality and an excellent tool for evaluating applicant competency. However, NUREG 1021, Section 201.D.2.a states that, " facility employees who had any direct involvement in training the license applicants shall not prepare the outlines for the written examinations or the operating tests." It further states that no one who provided 15% or more of the scheduled classroom instruction... may participate in developing the written examimtcn question When Region 111 examiners discovered the exam author was responsible ;or greater than 90% of the LSRO applicant's training, the examiners made a significant number of changes to the examination to ensure the examination process was not compromise The modified examination was then used to examine the LSRO applicant. Region 111 examiners determined that the examination author did not intend to circumvent any rules when writing the examination. Examiners determined that inadequate communications between the examination author and the NRC chief examiner concerning who could develop portions of the examination resulted in the improper examination submitta .2 Written Examination Examination Scoce The OCNPS training and operations departments were provided an opportunity to review the written examinations and provide comments prior to administering the examinations to the applicants, Observations and Findinas The OCNPS training staff reviewed the proposed NRC examinations snd submitted requests to change or replace several examination items. The requested changes incorporated ongoing revisions of the OCNPS operating procedures, and changed generic industry acronyms to plant specific acronyms used at QCNPS. The examiners reviewed the requested changes and, in most cases, made the change recommended by the plant staf All applicants passed the written examination with scores ranging from 85% to 89%. Conclusions

QCNPS training department personnel suggested changes that resulted in an l

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examination that was more plant specific than the proposed NRC examination and l incorporated recent changes made to the facility's operating procedures. Applicants appeared well prepared to take the writter. examinatio :

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O5.3 Administrative Job Performance Measures Examination Scooe The OCNPS training and operations departments were provided an opportunity to q review the administrative portion of the operating test and provide comments prior to l administering the test to the applicants. The administrative portion of the test was composed of four Job Performance Measures (JPMs) and two open reference 1 question ' Observations and Findinas 1 The OCNPS training staff reviewed the proposed NRC Admin JPMs for the SRO examination and submitted a request to replace one Admin JPM and make minor clerical changes to the other Admin JPMs. The staff review of the changes made by the examiners to the LSRO examination resulted in no staff comments. Examiners determined the requested changes to the SRO test would improve the overall quality of i the examination and incorporated the requested change )

Region ll1 examiners communicated the content of the replacement Admin JPM to QCNPS training department personnel and asked them to develop the replacement Admin JPM. A QCNPS trainer was assigned to develop the Admin JPM in accordance with the NRC communication. The trainer signed into his password protected hard drive on the OCNPS Local Area Network (LAN) computer and began work on the assignmen At the end of the day the trainer saved his work in a computer file and signed off the LAN computer. When the trainer returned to work and tried to access the computer file, the trainer was unable to locate the file. Since the trainer was unable to locate the file, and the trainer was using the LAN, the assumption was made that the JPM was uncontrolled and a different Admin JPM was developed to ensure examination integrit Subsequent to the development and use of the replacement JPM, the trainer located the missing file within his password protected area of the LA One Admin JPM involving tagout control was missed by both applicants. An Admin JPM with two errors (faults)in the JPM was administered to each applicant. Both applicants noted one of the two errors (an inadequate number of danger tags was returned with the tagout) but failed to detect the second error (positioner and verifier initials for one valve were by the same person). Conclusions Region 111 examiners determined that no examination compromised occurred and there was no intent on the part of the instructor to compromise the examination when the Admin JPM was temporarily lost. With the exception of the one Admin JPM it was apparent to the examiners that the applicants were well prepared for this portion of the examinatio .

f 05.4 Ooeratina Job Performance Measures Examination Scone The QCNPS training and operations departments were provided an opportunity to review the operating test and provide comments prior to administering the test to the applicants. The Operating JPM test was composed of ten JPMs, each of which had two follow-up question Observations and Findinas The QCNPS training staff reviewed the proposed JPMs and provided comments to enhance the cues that examiners provided to applicants as the applicants direct activities of operators outside the control room. The staff also provided comments to enhance the cues used while performing JPMs outside the control roo Instructors were careful to accurately prepare the simulator for each group of JPMs performed in the simulator which resulted in a smoothly coordinated administration of simulator JPM Conclusions The licensee suggested changes improved the overall clarity of the JPMs. The applicants appeared well prepared for this portion of the tes O5.5 Dynamic Simulator ScenarjQ1 Examination Scooe The OCNPS training and operations departments were provided an opportunity to review the operating test and provide comments prior to administering the test to the applicants. The Dynamic Simulator test was composed of two dynamic simulator scenario Observations and Findinas The staff members assisted the examiners with scenario development by suggesting malfunction ramp rates, types, and magnitudes. Suggestions made by the staff to improve the simulator scenarios during scenario validation resulted in an improved evaluation tool for determining applicant mastery of license dutie During administration of the dynamic simulator scenario test, examiners noted two occasions where applicants were complying with administrative procedure requirements that resulted in significant plant transients that may have been avoide (1) During one of the dynamic scenarios the SRO applicant was executing steps in the station's emergency operating procedures (QGAs) to protect primary

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containment. One QGA step stated that if operators are unable to maintain drywell temperature less than 280*F, they are required to emergency depressurize the reactor. An approved administrative procedure requirement modified the QGA requirement and stated that operators were to perform the emergency depressurization at a drywell temperature of 260*F. As drywell temperature reached 260*F the applicant stated he was going to emergency depressurize the reactor. The Balance of Plant (BOP) operator immediately reported he had initiated drywell sprays and that drywell temperature was decreasing. The SRO stated that he had already made the decision to emergency depressurize the plant and ordered the BOP operator to open safety relief valves to emergency depressurize the plant. When the order was given to emergency depressurize drywell temperature was 259'F and decreasing. The applicant's decision to emergency depressurize the reactor resulted in a significant transient on the reactor vessel, its associated piping and systems, and on the primary containment. Applying the rules of execution for EOP steps, the applicant did not need to order the emergency depressurization when the drywell j temperature tumed and started decreasing. The emergency depressurization j

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could have been avoide (2) A failure of the master feedwater level controller was conducted during one of the scenarios The controller failed such that it resulted in on overfeed conditio An SRO applicant at the Nuclear Station Operator (NSO) position took manual control of the controller and reduced its setpoint. The applicant balanced feedwater flow with steam flow and reported to the SRO that he had stopped the vessel level rise and had balanced steam and feedwater flows at a vessel level of +40 inches. The SRO then directed the NSO to scram the reactor based on ,

an administrative procedure that directed control room operators to sr tam the !

reactor anytime vessellevel reaches +40 inches. The NSO responded by 1'

manually scramming the reactor. The scram was ordered even though the parameter of concern was being controlled manually and was stable. The SRO applicant's decision to scram the reactor resulted in a significant, but avoidable, t transient on the reactor vessel and its associated piping and system I

Applicants normally used three-way closed loop communications during the test and only occasionally reverted to open loop communication A licensed operator was assigned to act as the BOP operator. The BOP operator executed his job function correctly without coaching the applicant The staff accepted the proposed changes to the operating LSRO test without commen c. Conclusions Communications practices of the applicants was generally good. Applicant compliance with administrative procedures resulted in two unnecessary, significant plant transient These two events suggested that a review of the station's administrative procedure concerning conduct of operations may be warrante i t

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05.6 Post Examination Activities The licensee informed the Chief Examiner by telephone that there were no post examination comments for the written examinatio One typographical error was found in the answer key. The key was changed to reflect the correct answer. Since both SRO applicants had correctly answered the question, their individual scores were increased by one poin O5.7 Simulator Fidelity Examiners observed one simulator modeling deficiency during the examination administration. A malfunction was inserted to fail the "A" Main Steam Line radiation monitor upscale. The malfunction did not produce the desired result on the monitor's associated recorder. All other portions cf the instrument failure responded as anticipated. A simulator problem report was generated to document the recorder failur The instrument failure did not impact the examination as a failure of the "C" Main Steam Line radiation monitor was inserted in the place of the "A" channel failure for the examinatio V. Management Meetings X1 Exit Meetina Summary The chief examiner presented the examination team's observations and findings to members of the licensee's management on March 20,1998. The licensee acknowledged the findings presented. No proprietary information was identified during the examination or at the exit meetin I

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PARTIAL LIST OF PERSONS CONTACTED Licensee l l A. Easley, Lead, Operations Requalification Training l C. Norton, Operations Staff Supervisor D. O'Rourke, BWR Training Supervisor L. Pearce, General Manager

M. Price, Lead Limited SRO License Training D. Snook, Lead, Initial License Training R. Svaleson, Operations Manager F. Tsakeres, Plant Training Manager

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NEC K. Walton, Resident inspector ITEMS OPENED, CLOSED, AND DISCUSSED Ooened i

None CJoled None Discussed

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O Enclosure 2 SIMULATION FACILITY REPORT Facility Licensee: Quad Citios Nuclear Power Station Facility Licensee Docket No: 50-254;50-265 Operating Tests Administered: March 16-18,1998 The following documents observations made by the NRC examination team during the January / February 1998, initiallicense examination. These observations do not constitute audit or inspection findings and are not, without further verification and review, indicative of non-compliance with 10 CFR 55.45(b). These observations do not affect NRC certification or approval of the simulation facility other than to provide information which may be used in future evaluations. No licensee action is required in response to these observation During the conduct of the simulator portion of the operating tests, the following items were observed:

ITEM DESCRIPTION

"A" MSL Rad Monitor When a faliure (high) of the "A" MSL rad monitor was inserted, the back panel recorder failed to read upscale. It continued to read the correct value for the plant condition i l

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SENIOR REACTOR OPERATOR Page 2

' ANSWER SHEET

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Multiple Choice . (Circle your choice) .

.-if you change your answer, write your selection in the blan a.b c d 022 a b c d 023 a b c - d

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002 ' a - b c d 003 . a b c d 024 a b c d 004 a - b c d - 025 a b c d 005 a bcd 026 a b c d 006 a b c d 027 a b c d 007 a b c d 028 a b c d 008 a b c d 029 a b c d 009 a - b c d 030 a b c d 010 - a b c d' 031 ' a b c d 011 abcq 032 a b c d 012 a b c d 033 a b ' c d 013 a b c d 034 a b - c d 014 a b c d 035 a b c d __

0.15 a b ~ c d 036 a b c d 016 a b c d 037 a b c d 017 a b c d 038 a b c d 018 a . b ' c d 039 a b c d -

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019 . a b c 'd 040 a b c d 020 a - b c d 041 abcd 021 abcd 042 a b c d l

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l SENIOR REACTOR OPERATOR Page 3 ANSWER SHEET Multiple Choice (Circle your choice)

If you change your answer, write your selection in the blan a b c d 064 a b c d 044 a b c d 065 a b c d 045 a b c d 066 a b c d 046 a b c d 067 a b c d M7 a b c d __ M8abcd 048 a b c d __ 069 a b c d 049 a b c d __ 070 a b c d 050 a b c d 071 abcd 051 abcd 072 a b c d 052 a b c d 073 a b c d 053 a b c d 074 a b c d 054 a b c d 075 a b c d 055 a b c d 076 a b c d 056 a b c d 077 a b c d 057 a b c d 078 a b c d 058 a b c d 079 a b c d 059 a b c d 080 a b c d 060 a b c d 081 abcd 061 a b~c d 082 a b c d 062 a b c d 083 a b c d 063 a b c d 084 a b - c d l

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SENIOR REACTOR OPERATOR . Page 4 ANSWER SHEET Multiple Choice (Circle your choice)

If you change your answer, write your selection in the blan a b c d l 088 a b c d 087 a b c d 088. a b c d 089 a b c d 090 a b c d 091 a b c d 092 a b c d 093 a b c d 094 a b c d -

095 a b c d 096 a b c d 097 a b c d 098 a b c d 099 a b c d 100 . a b c d i

(********" END OF EXAMINATION ""***"*)

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SENIOR REACTOR OPERATOR Page 5 Written Examination Guidelines (Read VerbatimJ 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 80.00 percent or greater. Every question is worth one poin . For an initial examination, the time limit for completing the examination is four hour For a requalification examination, the time limit for completing both sections of the examination is three hours. If both sections are administered in the simulator during a single three-hour period, you may retum to a section of the examination that was already completed or retain both sections of the examination until the allotted time has expire . You may bring pens and calculators into the examination room. Use only black ink or dark pencil to mark your answer .- Print your name in the blank provided on the examination cover sheet and the answer shee . Circle your answers on the answer sheets provided. Use only the answer sheets provided and do not write on the back side of the pages. If you decide to change your

' original answer, write the letter answer in the line following the letter selections. Any )

letter written on the line will be considered your final answe .- If the intent of a question is unclear, ask questions of the NRC examiner or the designated facility instructor only.

! Restroom trips are permitted, but only one applicant at a time will be allowed to legve.

l Avoid all contact with anyone outside t;m examination room to eliminate even the L appearance or possibility of cheating.

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l When you complete the examination,' assemble a package including the examination

! questions, examination aids, answer sheets, and scrap paper and give it to the NRC L examiner or proctor.- Remember to sign the statement on the examination cover sheet l indicating that the work is your own and that you have neither given nor received j

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assistance in completing the examination. The scrap paper will be disposed of

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immediately after the examinatio . After you have turned in your examination, leave the examination area as defined by the proctor or NRC examiner. If you are found in this area while the examination is still in progress, your license may be denied or revoke . Do you h!=ve any questions?

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- QUESTION: 001 (1.00)

Given the following conditions:

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Both Units are in Mode 4

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SBGT Train "A"in PRIMARY

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SBGT Train "B"in STANDBY

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System Engineering testing in progress

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A fault occurs on Bus 19 causing a loss of Bus 19 and RPS Bus "B". How will the SBGT system be affected?

l : SBGT system "A" 2-MO-7503 will receive an isolation signa SBGT system "A" will automatically start 25 seconds after the loss of RPS bus

"B".

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! SBGT. system "B" will automatically start 25 seconds after the loss of RPS bus

"B". SBGT system "A" and "B" initiation signals from Low-Low Reactor Water Level or Drywell High Pressure are defeate !

QUESTION:002 (1.00)

Unit Two is at 100% power. An error in surveillance testing has caused the "B" Core Spray .

auto initiation logic to become inoperable. Which ONE of the following is correct concoming

, Emergency Diesel Generator (EDG) operation? The Unit Two EDG will NOT auto start on a LOCA signal, The % EDG will NOT auto start on a LOCA signal from Unit Two ONL . The % EDG will NOT auto start on a LOCA signal from Unit One or Unit Tw The % EDG and the Unit Two EDG will auto start on a LOCA signal, but ONLY the Unit Two EDG output breaker will clos I e <

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. SENIOR REACTOR OPERATOR Page 7 QUESTlON: 003 (1.00)

Given:

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Unit One is in the refueling mode and core off-load is in progres RHR loop A' is in shutdown cooling with RHR pump A" runnin RHR loop B' tagged out of service.

l If reactor pressure increases to 105 psig and drywell radiation level increases to 20 R/Hr, which l ONE of the following describes the expected response of the RHR system?

ONLY the: MO-1001-20 and 21 will close.

l MO-1001-47 and 50 will close.

l l MO-1001-2gA,47, 50 will clos MO-1001-20,21. 29A,47, and 50 will clos QUESTION: 004 (1.00)

Which ONE of the following describes the operation of the Reactor Building to Suppression Chamber Vacuum Breaker inboard vacuum breaker valve? ASSUME the reactor building pressure is at a constant 14.5 psia.

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. The vacuum breaker valve automatically opens at a torus pressure of 14 psia, but will: close if torus pressure increases to 14.1 psia, and will fail closed if instrument air is lost, close if torus pressure increases to 14.1 psia, and will fail open if instrument air is los close if torus pressure decreases to 14 psia, and will fail open if instrument air is los close if torus pressure decreases to 14 psia, and will fail closed if instrument air is los i

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' SENIOR REACTOR OPERATOR Page 8 QUESTION: 005 (1.00) ,

With Unit One operating at full power, annunciator 901-3-C-14, TORUS VACUUM RELIEF VLV 20A NOT CLOSED, alarms. If this condition is confirmed to be true, what are the implications? Primary Containment integrity will be violated until the Torus to Reactor building l Vacuum Breakeris closed.

L i Drywell to Torus separation CANNOT be ensured until the Drywell to Torus Vacuum Breakeris close The check valve in the Torus to Reactor Building Vacuum Breaker line is now providing Primary Containment integrity.

l The check valve in the Drywell to Torus Vacuum Breaker line is now providing

! separation between the Drywell and Torus.

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l QUESTION: 006 (1.00)

GIVEN:

h - Rx Power; 40%

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Rx Water Level: +42 inches and stable

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Rx Pressure: 815 psig and decreasing

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No operator actions have been take .What automatic actions should have occurred? Group One isolation and Reactor $ cram Main Turbine Trip, HPCI Turbine Trip, and Reactor Scram Main Turbine Trip, HPCI Turbine Trip, and Group One isolatio Main Turbine Trip, HPCI Turbine Trip, Reactor Scram and Group One isolatio .

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SENIOR REACTOR OPERATOR Page 9 QUESTION: 007 (1.00)--

SELECT the set of parameters that are needed to satisfy all Standby Liquid Control (SBLC)

System Technical Specification OPERABILITY requirements:

SBLC System Sodium Sodium l Tank Volume Pentaborate Pentaborate l (gallons) Solution Solution Temperature

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(weight %) (degrees Fahrenheit)

l ,700 1 ,750 1 J l .0 75 .0 75

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QUESTION: 008 (1.00) I A reactor startup is in progress. The reactor is subcritical with all IRMs on range 4, except for ,

IRMs 14 and 18 which are bypassed. The following IRM readings are observed: j IRM channel 1148/125 of scale IRM channel 12 52/125 of scale ,

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IRM channel 13 49/125 of scale  !

IRM Channel 15 55/125 of scale l IRM Channel 16 58/125 of scale IRM Channel 17 73/125 of scale j If all IRM Range Switches were placed to Range 5, what response would be expected? j No protective function ! A Control Rod Out Block onl l A Control Rod Out Block and a half Scram, i l A Control Rod block and a Full Reactor Scra l i

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SENIOR REACTOR OPERATOR Page 10 l L QUESTION: 000 (1.00)

Which ONE of the following will cause the Yarway Narrow Range Level indication to indicate lower than actual?

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) Bellows lea ! Equalizing valve lea Variable leg sensing line leak.

' Boil off from the reference le QUESTlON: 010 (1.00)

Given the following conditions on Unit One:

Drywell Pressure 3.5 psig steady RPV Level -60 inches slowly lowing ADS in inhibit RPV Pressure was reduced to 100 psig by manually cycling Safety Relief Valve RHR is running in the LPCI mode with the MO-1001-28 and MO-1001-29 ope How will the LPCI mode of RHR respond if RPV pressure rises to 350 psig? ONLY the MO-1001-28 will auto clos ONLY the MO-1001-29 will auto clos Both the MO-1001-28 and MO-1001-29 will auto clos Both the MO-1001-28 and MO-1001-29 will remain ope l l

SENIOR REACTOR OPERATOR Page 11

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QUESTION: 011 (1.00)

A high drywell pressure signal (2.5 psig) coincident with a low low reactor level signal (-59") is received but only one Core Spray pump and NO RHR pumps start. ADS will initiate: in 110 second in 8.5 minute seconds after an RHR pump is starte when the second CS pump or any RHR pump is started.

QUESTION: 012 (1.00)-

Given the Following conditions on UNIT TWO:

- The Unit is in MODE 5

- Shutdown Cooling in service

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Both Recire Pumps are OFF

- MSIVs closed if Shutdown Cooling (SDC) is lost due to a fault in the SDC logic which of the following would be an indication that Temperature Stratification exists? An unexpected increase in reactor pressure, An observed increase in reactor metal temperatures which resulted from a decrease in reactor vessellevel, An observed increase in reactor metal temperatures which resulted from an increase in reactor water temperature:. ,

I An observed increase in reactor vessel shell temperatures which resulted from l an increase in reactor vesselleve .

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SENIOR REACTOR OPERATOR Page 12 QUESTION: 013 (1.00)

Which ONE of the following conditions requires entry into QGA 300, SECONDARY CONTAINMENT CONTROL 7 " of water on the floor below the toru Reactor Building Differential Pressure below 0". Reactor Building ventilation exhaust radiation level of 2 mR/h Reactor Building floor drain sump pump inoperable with sump level below the high level alarm point, but increasing.

QUESTlON: 014 (1.00)

The MAX SAFE values for temperature, radiation, and water level in the Reactor Building for QGA 300, " SECONDARY CONTAINMENT CONTROL," are based on the maximum value(s): at which NO equipment will fai expected to be seen during an acciden expected to be seen during normal operation at which equipment needed for safe shutdown of the plant will NOT fai ,

SENIOR REACTOR OPERATOR Page 13 QUESTION: 015 (1.00)

During post LOCA conditions, with drywell temperature at 260*F, and reactor building temperature at 198'F, the following reactor water levels are noted:

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Yarway wide range -110 inches

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GEMAC lower 400 -140 inches

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Yarway narrow range -50 inches

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GEMAC upper 400 +80 inches Which of the above level indicators CANNOT be used in these plant conditions? GEMAC lower 400 GEMAC upper 400 Yarway wide range Yarway narrow range QUESTION: 016 (1.00)

During Non-ATWS RPV Flooding, if only 4 ADS valves open and a Feedwater pump is available, why are the MSIVs, MSL drains, and RCIC steam isolation valves closed? If a feedwater pump is available, it can be used to flood the RPV irrespective of whether the RPV is vented or no These valves are required to be closed to ensure the proper flooding pressure can be reached using a feedwater pum These valves are required to be closed to ensure the proper RPV flooding level can be reached using a feedwater pump, If a feedwater pump is available it can provide sufficient feed flow to overcome any postulated LOCA, and thereby flood the cor i l

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iSENIOR REACTOR OPERATOR Page 14 QUESTlON:017 (1.00)

There are two divisions of ARI valves installed tn the scram air header. Each division of valves consists of: One ARI valve that opens to depressurize the scram air header. One three-way ARl valve that closes off the supply air and vents the scram air header to atmospher Two ARI valves that open to depressurize the scram air header. One three-way ARI valve that closes off the supply air and vents the scram air header to j atmosphere,  ! One ARI valve that opens to depressurize the scram air header. Two three-way ARI valves that close off the supply air and vents the scram air header to atmospher Two ARI valves than open to depressurize the scram air header. Two three-way ARl valves that close off the supply air and vents the scram air header to atmospher QUESTION: 018 (1.00)

Which of the following combinations stctes the correct power supplies to the two solenoids associated with each Main Steam isolation Valve (MSIV)?

INBOARD MSIVs OUTBOARD MSIVs VDC "A" 125 VDC "B" RPS Channel"A" RPS Channel"B" VDC "A" 125 VDC "B" RPS Channel "B" RPS channel"A" VDC "B" 125 VDC "A" RPS Channel"B" RPS Channel "A" VDC "B" 125 VDC "A" 1 RPS channel"A" RPS Channel"B" l l

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SENIOR REACTOR OPERATOR Page 15 QUESTION: 019 (1.00)

Given the following initial conditions:

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Reactor Power 100 %

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Load Limit 100 %

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Max. Combined Flow 110 %

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EHC Pressure Set 920 psig

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Steam Throttle Pressure 950 psig Which ONE of the following describes the correct tuttine response for the given action? Reducing the load limit potentiometer to 92% would result in a control valve demand of 92% and a bypass valve demand of 8%. Reducing the load limit potentiometer to 92% would result in a control valve demand of 100% and a bypass valve demand of 8%. Reducing the max. combine flow potentiometer to 90% would result in a control valve demand of 90% and a bypass valve demand of 0%. Reducing the max. combine flow potentiometer to 90% would result in a control

valve demand of 90% and a bypass valve demand of 10%.

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QUESTION: 020 (1.00)

While operating HPCI for level or pressure control in accordance with the QGAs, you are directed to defeat the HPCI high torus level transfer in accordance with QCOP 2300-9. Which .

ONE of the following describes the effect this has on the HPCI suction valve logic?

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The HPCI suction valves: will automatically transfer if a CCST low level signal exist will automatically transfer if a torus high level signal exist can only be transferred manually if a CCST low level signal and a torus high ,

level signal exis j can NOT be transferred manually or automatically on a CCST low level signal or

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QUESTION: 021 (1.00)

A plant startup is in progress on Unit One. All prerequisites are met for rolling the main turbine, FAST STARTUP RATE has been selected. The ANSO presses 1800 RPM on SPEED SET RPM and verifies that the MAIN STOP VLV #2 opens to 100%, then MAIN STOP VLVs #1, j

1. 3,and #4 ramp ope Which ONE of the following describes the correct opening sequence for the combined intermediate stop valves (CIVs)? , 2, and 3 open to 100%, then 4, 5, and 6 ramp ope ,3, and 5 open to 100%, then 2,4, and 6 ramp ope ,3, and 4 open to 100%, then 2,5, and 6 ramp ope , 2, and 5 open to 100%, then 3,4, and 6 ramp ope I i

QUESTION: 022 (1.00)

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' Given the following conditions: j

- Mode Switch in RUN

- APRM 3 indicates Downscale 1

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APRM 4 indicates Downscale l

- IRM 14 is inop and Bypassed

- IRM 17 is Inop and Bypassed  ;

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IRM 18 indicates Hi HI 1 l

Which ONE of the following describes the trip signals generated by the given conditions?

(Assume no operator actions.) Alarms only Alarms and a Rod Block only Alarms, Rod Block, and a half Scra Alarms, Rod Block, and a full Scram.

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SENIOR REACTOR OPERATOR Page 17 i

QUESTION: 023 (1.00)

Which of the following is the function of the IRM MANUAL BYPASS JOYSTICKS? The j Joysticks allow bypassing:  ! of 4 IRM channels on RPS bus A AND 1 of 4 channels on RPS bus of 4 IRM channels on RPS bus A OR 1 of 4 channels on RPS bus B.

I of 4 IRM channels on RPS bus A AND 2 of 4 channels on RPS bus of 4 IRM channels on RPS bus A OR 2 of 4 channels on RPS bus QuedTION: 024 (1.00)

Give:

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A reactor startup is in progress, the reactor has been made critical 3

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All IRMs are reading approximately 45 on range 2

- The SRMs have been partially withdrawn from their full in position (all of the "IN" lights are OFF) to maintain their indication between 100 and 1.0E5 If SRM 22 failed downscale, which of the following best describes the plant response to this failure? (All shorting links are INSTALLED).

I ONLY alarms will occu l ONLY a Rod Block and alarms will occu A full SCRAM, a Rod Block and alarms will occu , ONLY a half SCRAM, a Rod Block and alarms will occu ;

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l QUESTION: 025 (1.00)

Unit Two was at 100% power when a Main Steam Line high-high radiation alarm was received and confirmed.' in accordance with QCOA 1700-5, ABNORMAL MAIN STEAM LINE RADIATION, which ONE of the following is a correct IMMEDIATE OPERATOR ACTION?

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Place control switch for AO-2 ...

HOLDUP PIPE DRN to CLOS b. - 5408 OFF GAS FILT DRN to CLOS PRESS DRN TK DISCH VALVE to CLOSE.

! OG DISCH TO STACK or VENT to CLOS QUESTION: 026 (1.00)

The following conditions exist for unit The reactor has scrammed and the mode switch is in SHUTDOW The problem that caused the scram has been identified and correcte Annunciator " CHANNEL A/B DISCH VOLUME HIGH LEVEL"is sealed i Which ONE of the following describes the reactor protection system (RPS) response when you

! place the Scram Discharge Volume High Level Keylock switch in BYPASS, followed by placing the scram reset switch in both directions, and finally placing the mode switch in STARTUP7 The RPS will reset and remain rese The RPS will reset and again scra Nothing will occur due to the present plant condition The RPS will reset when the scram discharge volume drains.

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SENIOR REACTOR OPERATOR Page 19 QUESTION: 027 (1.00)

Unit One is operating at 25% power. MSIV 1-203-1A has a faulty limit switch that is generating a less than 90% open signal to RPS. Which ONE of the following MSIVs, if closed, will cause a Half Scram on the B' RPS Channel? A C D QUESTION: 028 (1.00)

With the plant at 80% power and rod H-8 selected, APRM 3 failed downscale. What effect will this have on the Rod Block Monitor system and what can be done to correct the problem? RBM 7 will initiate a Rod Out Block. To correct, bypass the RBM using the RBM Bypass Joy Stic RBM 8 will initiate a Rod Out Block. To correct, bypass the RBM using the RBM Bypass Joy Stic RBM 7 will be automatically bypassed. To correct, bypass APRM 3 usir eg the APRM Bypass Joy Stick to change the reference APRM to APRM RBM 8 will be automatically bypassed. To correct, bypass APRM 3 using the APRM Bypass Joy Stick to change the reference APRM to APRM . _ _ _ _ _ - _ _ _ _ _ -

SENIOR REACTOR OPERATOR Page 20 QUESTION: 029 (1.00)

Which ONE of the following statements describes the effect on the RWM of placing the RWM system in BYPASS? The RWM continues its rod scanning functio The RWM annunciator is bypassed while in this mode.

i The RWM is functionally removed from the RBM syste The downloading of a new sequence while in this mode IS allowe QUESTION: 030 (1.00)

A double-ended shear of a reactor recirculation suction line near the vessel at full power occurs (DBA LOCA). Before an ECCS system injects to the core, how far will reactor vessel level drop? To the top of active fue To approximately 1/3 core heigh To approximately 2/3 core height.

, Below the bottom of the active fue QUESTION: 031 (1.00)

During an ATWS, QGA 101 directs you to place ADS inhibit switch in INHIBIT. Which of the following states the reason for this requirement? ADS actuation would result in the removal of boron after it has been injecte Core damage could .rame.4 from a large power excursion if low pressure ECCS systems were to injec ADS System flow rate is incapable of assuring adequate core cooling through steaming above 5% reactor powe ,

1 ADS actuation will cause a loss of core cooling and subsequent core damage which may have otherwise been avoide J

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SENIOR REACTOR OPERATOR Page 21 QUESTION: 032 (1.00)  ;

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l As stated in Control Room Access procedure, Which of the following personnel have completely unrestricted access to the Control Room?

l Q and SA. representatives performing overview functions Maintenance Superintendent Services Superintendent l NRC personnel Operations Manager Shift Operations Supervisor , 2, & 4 l , 5 & 6 , 5 & 6 ,4 & 5 I

QUESTION: 033 (1.00)

What is the purpose of a caution card? To prevent operation of equipmen ,

l b.- To protect personnel working on equipmen ! To identify a problem that requires a work reques )

d. . To inform people of temporary changes in the status of equipmen i i

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SENIOR REACTOR OPERATOR Page 22 I

. QUESTION: 034 (1.00)

You have been called out to work 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of overtime. You have worked 64 hours7.407407e-4 days <br />0.0178 hours <br />1.058201e-4 weeks <br />2.4352e-5 months <br /> in the last seven days which includes 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> yesterday. You have had 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> off and this is your RD Can you work the overtime and stay ;,;ihiri the OT guidelines? Why or why not ? No, you can only work 64 hours7.407407e-4 days <br />0.0178 hours <br />1.058201e-4 weeks <br />2.4352e-5 months <br /> in 7 day No, you can only work 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in 48 and 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> in every 3 Yes, ycu can work anytime management calls you, it is their responsibility to verify you are eligibl Yes, you can work up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in 7 days and 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> every 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> if you've had at least 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of QUESTION: 035 (1.00)

As required by 10CFR26, " Fitness for Duty Programs," which ONE of the following is the MINIMUM time an operator must abstain from the consumption of alcohol prior to any SCHEDULED shift? hours hours hours hours

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SENIOR REACTOR OPERATOR Page 23 QUESTlON: 036 (1.00)

In accordance with QCAP 0630-06 EXPOSURE AUTHORIZATION & CONTROL, what is the MAXIMUM allowable TEDE when entering an area to prevent conditions which could cause injury to other people? rem rem rem rem QUESTION: 037 (1.00)

The DGCWP is in a normal standby condition. The 2B FPCWP is running. The DGCWP/FPCWP Feed selector switch is placed in the DGCWP position. Predict the effect on !

the diesel cooling water pump and fuel pool cooling water pum The DGCWP will start and the 2B FPCWP will shut of The DGCWP will NOT start and the 2B FPCWP will shut off, The DGCWP will start and the 2B FPCWP will continue to ru The DGCWP will NOT start and the 2B FPCWP will continue to ru )

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SENIOR REACTOR OPERATOR Page 24 l

QUESTION: 038 (1,00)

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You are restoring reactor building ventilation after an isolation. The signal that caused the isolation has cleared and has been reset. The damper control switches at the 912-5 panel are still in the NORMA 1.-AFTER-OPEN position. What is the response of the reactor building vent isolation dampers when the reactor building vent isolation reset pushbutton is pressed on the 912-5 panel? The dampers will reope No response. The dampers reopened when the signal that caused the isolation !'

was rese The dampers will NOT reopen until the reset pushbutton at the 2251(2)-24X ,

i panelis depressed also, The dampers will NOT reopen until the isolation damper control switches are taken to CLOSE and back to OPE i

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QUESTION: 039 (1.00)

If the Unit 1 NSO is directing a surveillance from the control room and a step requires the non-licensed operator to operate a valve locally, how would this be recorded on the surveillance sheet? The NSO must initial the step for the non-licensed operato The non-licensed operator must initial the sheet after that step is complet The non-licensed operator must initial the sheet after the surveillance is complet The NSO must put the non-licensed operators initials in the space, followed by his/her own initial SENIOR REACTOR OPERATOR Page 25 OUESTION: 040 (1.00)

Which One of the following is a control room immediate action required by QOA 0010-05, " Plant Operations with the Control Room inaccessible?" Place the Mode Switch in SHUTDOW Close the MSIVs from the 901(2)-3 pane Trip the main turbine from the 901(2)-7 pane Manually scram the reactor from the 901(2)-5 pane )

QUESTION: 041 (1.00)

With the 1/2C RBCCW Pump running on Bus 19, what actions must be taken to restart the pump on Bus 19 if a LOCA on Unit One causes Drywell pressure to increase above 2.5 psig? ;

I Place BOTH of the U1 DIV I and DIV 11 DW CLR/RBCCW/FPC TRIP BYPASS Switches in the BYPASS position and manually restart the pum Place BOTH of the U1 DIV I and DIV ll DW CLR/RBCCW/FPC TRIP BYPASS Switches in the BYPASS position and the pump will restart automaticall Place EITHER the U1 DIV I cr the U2 DIV 1 DW CLR/RBCCW/FPC TRIP BYPASS Switches in the BYPASS position and manually restart the pum Place EITHER the U1 DIV I or the U2 DIV 1 DW CLR/RBCCW/FPC TRIP BYPASS Switches in the BYPASS position and the pump will restart automatically, i

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QUESTION: 042 (1.00)

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The bypass valves are being used to lower reactor pressure. Concurrently, condenser vacuum is slowly decreasing. Which one of the following vacuum readings corresponds to the lowest

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value at which the bypass valves will still be effective in reducing reactor pressure? (Consider ONLY actual plant setpoints per QOA 3300-02 LOSS OF CONDENSER VACUUM for your answer) O inches Hg vacuum ' inches Hg vacuum inches Hg vacuum inches Hg vacuum QUESTION: 043 (1.00)

Which of the following are the AUTOMATIC ACTIONS associated with the loss of 120/240 VAC ESSENTIAL SERVICE BUS? ) half Reactor Scram and half Group One isolatio Recirculation Sample and TIP Ball valves auto clos Reactor Building and Control Building ventilation isolate SBGTS auto starts and Reactor Building ventilation isolates.

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SENIOR REACTOR OPERATOR Page 27 QUESTION: 044 (1.00)

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Given a total loss of UNIT 2125 VDC supply, which of the following is correct conceming the HPCI logic and relief valve controllers? Unit 2 HPCI logic transfers to Unit 1 125 VDC. Unit 1 relief valve controllers transfer to Unit 1125 VD Unit 1 HPCI logic transfers to Unit 1 125 VDC. Unit i relief valve controllers transfer to Unit 1125 VD I Unit 2 HPCI logic transfers to Unit 1 125 VDC. Unit 2 relief valve controllers l transfer to Unit 1125 VD ,

I Unit 2 HPCI logic transfers to Unit 1 125 VDC. Unit i relief valve controllers transfer to Unit 2125 VD I l

' QUESTION: 045 (1.00)

The High Reactor Water t.evel Trip at 48 inches is designed to protect the: Main Turbine from moisture carryove MSIVs from excessive hydraulic loading.

! Reactor feed pumps from run-out conditions, Main Turbine Bypass and Control Valves from thermal shoc '

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QUESTION: 046 (1.00)

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While operating at power the NSO reports he has noticed a sudden rise in total core flow and l suspects jet pump failure. To confirm jet pump failure you should direct the NSO to look for a: rise in core thermal powe !

l rise in main generator electrical outpu drop in core plate differential pressur i drop in individual recire pump flow for a given spee !

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SENIOR REACTOR OPERATOR Page 28 QUESTION: 047 (1.00)

The reactor has been scrammed and the Mode Switch taken to SHUTDOWN in response to an instrument air header rupture that has resulted in a loss of Instrument Air on Unit 2. Which one l of the following describes how the operation of the MSIVs will be affected by this condition?

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i The inboard MSIVs would remain open; the outboard MSIVs would close.

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l All MSIVs would remain open since the MSIV Instrument Air Cross-tie will l automatically open.

i f All MSIVs would remain open since the drywell pneumatic system will l automatically align to supply the MSIV The inboard MSIVs would close when their accumulators discharged; the outboard MSIVs would remain open.

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.I QUESTION: 048 (1.00)

Which of the following is the maximum RPV pressure to maintain in order to be assured the Heat Capacity Limit will NEVER be exceeded? psig psig psig psig i

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QUESTION:049 (1.00)

l Unit 1 was opercting at 75% power when both CRD pumps failed. In accordance with QCOA 0300-01 CONTROL ROD DRIVE PUMP FAILURE, which of the following conditions would

! require the Mode Switch to be placed in SHUTDOWN? When the first accumulator trouble alarm is receive When the second accumulator trouble alarm is receive When the third accumulator trouble alarm is receive When fourth accumulator trouble alarm is receive I

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QUESTION: 050 (1.00)

Given:

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Unit 2 was operating at 100% power ,

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A leak developed on the recirc system  !

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Drywell pressure to increase to 3.0 psig.

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- The reactor scrammed, the UNSO placed the Rx Mode Switch in SHUTDOWN

- Reactor water level decreased to + 5" inches before being restored to the normal ban !

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Which ONE of the following Primary Containment isolation System group isolations should have

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occurred? i Group i ONL Group ll ONL Group 11 and lli ONL Group I and ll ONLY.

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l l-SENIOR REACTOR OPERATOR Page 30 QUESTION: 051 (1.00)

Which of the fo, lowing is the minimum Emergency Plan classification for entry into QGA 400 RADIOACTIVITY RELEASE CONTROL 7 Unusual Event -

Alert Site Area Emergency j General Emergency i l

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QUESTION: 052 (1.00)

l- Which of the following is the ADMINISTRATIVE daily exposure limit?

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! mrem DDE mrem DDE mrem DDE l r- mrem DDE l

i QUESTION: 053 (1.00)

i Which of the following describes the operation of motor operated (MO) throttle valves? ! Red control switches indicate that the valve throttles both open and shut.

l Red control switches indicate that the valve throttles shut and seals in open.

! Yellow control switches indicates that the valve throttles open and seals in shu , All MO throttle valves should be driven closed for a minimum for 25 seconds I

after receiving closed indicatio !

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l l SENIOR REACTOR OPERATOR Page 31 QUESTION: 054 (1.00)

Which of the following describes the method used to take station equipment out-of-service (OOS).

, Daughter OOSs may rely on a Stand Alone OOS for configuration only.

l Daughter OOSs may NOT be taken OOS until the associated Stand Alone OOS

, is taken OOS.

t .. Stand Alone OOSs may relay on a Parent OOS for configuration control but has its own Zone Of Protectio Parent OOSs are taken by the Operations Department to establish a desired plant configuration on a specific system or component.

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' QUESTION: 055 (1.00)

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, in accordance with QCAP 0230-05, which of the following is the PREFERRED method to perform an independent verification on a manually throttled valves? Perform a second valve operation to verify the positio Observe the initial operator's action in positioning the throttled valv Observe flow indication through the throttled valve's system during system lineu Perform an independent visual check of the valve position by comparing the actual valve position with the required valve position.

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SENIOR REACTOR OPERATOR Page 32 QUESTION: 056 (1.00)

Which ONE of the following describes the operation of the PCIS Status Box on the Safety l Parameter Display System (SPDS)? A half Group Isolation signal with no valve movement will cause a RED ligh A full Group Isolation signal with no valve movement will cause a GREEN light.

! A full Group Isolation signal with 2 out of 2 isolation valves in a line closed with cause a RED ligh I A full Group Isolation signal with 1 out of 2 isolation valves in a line closed will cause a GREEN ligh QUESTION: 057 (1.00)

You are operating a MOV electrically from the local control station. What limits are in effect for the number of starts you can attempt?

l An MOV is limited to five starts within one minute followed by a thirty minute cool

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off perio An MOV is limited to one start per minute for five start attempts followed by an hour cool off perio An unlimited number of starts may be attempted for one minute then a fifteen minute cool off period is require ! An MOV is normally limited to two start attempts unless permission is given by the system engineer to continu '

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SENIOR REACTOR OPERATOR Page 33 QUESTION: 058 (1.00)

How are instruments and equipment highlighted to indicate they have significance during Emergency Operating Procedure (QGA) execution? Control room instruments which are related to entry conditions have black dots, Valves which are infrequently used and require quick identification have yellow label Alarms which indicate that a QGA entry condition has been reached has a red borde Alarms which indicate that a QGA entry condition is being approached has a red borde QUESTION: 059 (1.00)

Given:

- Drywell flooding is in progres Containment water level is 22 feet and risin Two core spray pumps and all four LPCI pumps are injecting with suction from the toru Standby Coolant is injectin RPV pressure is 40 psig and risin ADS valves are open, the others are stuck shu Torus pressure is 44 psig and risin The torus is being lined up to vent through the APC You should: stop injecting with all system stop injecting with Standby Coolan swap both Core Spray pump suctions to the CCS swap ONLY one loop of LPCI pump suction to the CCS !

SENIOR REACTOR OPERATOR Page 34 QUESTION: 060 (1.00)

Given:

Time 0730 - Reactor scram and all rods inserted to 0 Time 1110 - Saturation conditions are reached in the Drywel Time 1200 -

All 5 ADS valves are open with RPV pressure greater than or equal to 58 psig above torus pressur Time 1345, -- The Drywell has been successfully sprayed and is no longer saturate The IMs have backfilled reference legs on the water level indicator It is now 1400. What is the MAXIMUM time that injection can be stopped and water level lowered to observe water level indicator response? minutes minutes minute minute QUESTION: 061 (1.00)

Given certain ATWS conditions, why does QGA 101 direct the lowering of RPV water level to ,

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-142 inches? This ensures natural circulation is maintained.- This is done to reduce reactor power to protect primary containment integrit )

i = This corresponds to the lowest RPV level at which fuel failure will NOT occu i l

' This maintains indicated reactor vessel water level above two-thirds core height to ensure containment cooling remains activ !

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1 SENIOR REACTOR OPERATOR Page 35 QUESTION: 062 (1.00)

An electrical fault has resulted in the following condition Partialloss of Drywell coolin Drywell pressure 5.0 psig slowly rising.

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Drywell temperature 285*F and slowly rising.

l l What are the required operator actions in accordance with the Primary Containment Control flowchart?

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l Spray the drywell.

! Scram and continue in QGA 20 ' Continue in QGA 200 and enter QGA 300.

l Scram, and enter QGA 1(,0 and QGA 500- QUESTION: 063 (1.00)

Why are the ADS valve control switches left in manual when performing an RPV blowdown? All valves will be fuii ope To ensure the RPV remains depressurize If the switche: . were retumed to off, the ADS valves would only open cn setpoint pressur i To allow the valves to cycle on their automatic pressure setpoint following the initial blowdow QUESTION: 064 (1.00)

Where does the Unit Supervisor report when the Control Room is evacuated? TSC Aux Electric Room Reactor Building Second Floor l L Reactor Building Third Floor

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l QUESTION: 065 (1.00)

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Unit One is operating at 100% power. The1A Gore Spray pump is INOPERABLE. If the %

diesel generator becomes INOPERABLE, determine what action (s), is(are) now require Enter a 14 day LCO for Core Spray, Enter a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> LCO for the Core Spra Place Unit One in HOT SHUTDOWN within 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Enter a 7 day LCO for Core Spray and for the Diesel Generator l

l QUESTION: 066 (1.00) .

l The following Unit 2 plant conditions are given: l

- Reactor power is 40%

- QCOS 1300-5, " Quarterly RCIC Pump Operability Test" is in progress

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Torus water temperature is 73*F

- Torus water temperature is increasing at 4*F every 22 minutes j

- Torus cooling is in service l What is the MAXIMUM amount of time this test may continue without Tech Specs requiring the I test to be terminated?  ! minutes minutes minutes 1 minutes l

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SENIOR REACTOR OPERATOR Page 37 QUESTION: 067 (1.00)

Given the following plant conditions:

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Unit 2 is operating at 40% powe The "B" Recirculating pump is operating with a core flow of 40 mlbm/h The "A" Recirculating pump is being restarted after a spurious trip and is currently running at minimum spee The discharge valve for the "A" Recirculating pump (SA) has been opened to 85% ope A fault in the valve control circuit causes the discharge valve (SA) to move in the

'( closed direction until fully close .Which of the following actions describes the immediate recirculation system response with no operator action? The "A" pump will continue to run at minimum spee The "A" pump will trip when the valve is less than 90% ope The "B" pump will run back to 32% speed due to excessive flow mismatc The "A" pump will trip when the scoop tube lockup occurs due to excessive speed mismatc QUESTION: 068 (1.00)

Given: _

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Core thermal power is 65% of rate Total core flow is 50 mlbm/h Which ONE of the following will drive the plant closer to the instability region? Inserting control rod Raising reactor water leve Raising feed water temperatur Reducing reactor recirculation pump flo _ _ _ .

SENIOR REACTOR OPERATOR Page 38 l

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! QUESTION: 069 (1.00)

l The plant was operating at 100% rated power when the "A" recirculation pump tripped. As an immediate action of QCOA 0202-4, the operator is directed to monitor for oscillations indicating l core instability. Which of the following is an indication that core wide instabilities are occurring? Regular oscillations of reactor water level with a 2-3 second periodicit Excessive core plate d/p noise exceeding a value of 0.5 psi peak to peak.

l Oscillations on the LPRMs with a characteristic periodicity of 1.5 to 2.5 seconds, High values of APRM noise tha' ,,. cur with no regular frequency and are

random in magnitude.

, OUESTION: 070 (1.00)

Unit One is Shutdown in MODE tlE is a} l00"7o pou>e < hCon4i e 4D'f" *

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Core Offload is in progres d ( 0 0 1o f # u d o ulg & li [ f e t !

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All systems are currently operable *4 *

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A lightning strike causes a loss of Transformer 12.

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The % EDG fails to pick up bus 13- ;

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- The ANSO cross ties bus 13-1 to 23-1 and reenergizes all required loads from j l

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Determine what actions, if any, are now require No actions are required at this tim Suspend CORE ALTERATIONS and suspend handling of irradiated fuel in the secondary containment.

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l CORE ALTERATIONS may continue, suspend handling of irradiated fuelin the

! secondary containmen < Suspend CORE ALTERATIONS, handling of irradiated fuel in the secondary containment may continu SENIOR REACTOR OPERATOR Page 39 QUESTION: 071 (1.00)

Which two of the below listed components are a part of the definition of an intact SECONDARY

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l CONTAINMENT? At Least ONb door in each access opening is close . The Reactor Building vent stack is operabl . .

The Standby Gas Treatment System is operabl . The drywell airiock is operable or locked close . All automatic isolation valves / dampers are operable or locked close and 3 and 4 and 4

, and 5 QUESTION: 072 (1.00)

Which of the following is NOT a design basis for the Control Room HVAC system? Provide HVAC to the auxiliary electric equipment roo To detect and limit the amount of smoke in the Control Roora atmosphere, To detect and limit the amount of Ammonia in the Control Room atmospher , To detect and limit the amount of Carbon Monoxide in the Control Room atmospher I J

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SENIOR REACTOR OPERATOR . Page 40 QUESTION: 073 (1.00)

L With the reactor operating normally at full power, RWCU valve MO-2-1201-133, DEMIN BYPASS, is being throttled to maintain RWCU pump discharge pressure while one RWCU filter domineralizer is removed from service. During the evolution the bypass valve is opened too far, causing RWCU pump discarge pressure to drop below 1050 psic. Which of the following actions may occur? RWCU pumps will experience runou RWCU pumps will trip on low pump flo RWCU system will isolate on high temperatur The remaining on-service RWCU domin will isolate on low flo QUESTION: 074 (1.0u) .

The function of the Containment Cooling 2/3 Level & ECCS Initiation Bypass Keylock switch in the " MANUAL OVERRIDE"in te: Allow Suppression Pool Spray to be used even if dywell pressure is less than 1 psi b.' Allow Drywell and Suppression Pool Spray to be used even if reactor vessel level is below -191".

. Allow Drywell Spray to be used even if a LPCI initiation signal is present and vessel level has reached the Top of Active Fuel (TAF). Allow Suppression Pool Cooling to be used even if a LPCI initiation signal is present and vessel level has reached a level of 2/3 normal leve .

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SENIOR REACTOR OPERATOR Page 41 l

QUESTION: 075 (1.00)

l Unit One is operating at 100% power. The "B" SJAE Rad Monitor is inoperable and failed UPSCALE. The "A" SJAE fails DOWNSCALE. What indications would be expected on the SJAE Red Monitor interval timer (Off Gas 15 minute Timer)? Both lights would be ON.

l- Both lights would be OF The right light would be OFF. The left light would be O The right light would be ON. The left light would be OF QUESTION: 076 (1.00)

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Level I Risk Activities are approved by the:

l l Work Control Superintendent Operations Manager

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l Station Manager Site Vice President QUESTION: 077 (1.00)

Which ONE of the following is a Design Basis of HPCl? Must be able to function independent of on-site power source .Make up water to the vessel in the event of a loss of coolant situation that does NOT result in rapid vessel depressurizat;o Assures that the reactor core is adequately cooled to limit fuel clad temperature in the event of a large break in the reactor coolant syste i Assures that the reactor core is adequately cooled to limit primary containment )

. pressure in the event of a small break in the reactor coolant syste l

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SENIOR REACTOR OPERATOR Page 42

' QUESTION: 078 (1.00)

' Given that the Reactor Mode switch is in REFUEL, which ONE of the following will cause a refueling platform Main Hoist Motion Block? NOT all rods in (only) NOT all rods in and the Main Hoist loaded NOT all rods in and the Bridge near or over the reactor vesse NOT all rods in, the Bridge near or over the reactor vessel, and the Main Hoist loade QUESTION: 079- (1.00)

While moving a spent fuel bundle in the Fuel Pool, a Fuel Pool Storage Low Level Alarm is received and Fuel Pool Level is confirmed to be decreasing. Which ONE of the following is the expected operator action IAW QCOA 1900-01 LOSS OF WATER LEVEL IN THE FUEL STORAGE POOL OR REACTOR CAVITY? Retum bundle to its original location, Suspend bundle movement where it i Place bundle in the nearest storage locatio Lower bundle as far as possible without moving refueling bridg ,

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l QUESTlON: 080. (1.00)

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Which of the following is a responsibility of the Unit Supervisor during refueling? The authority to halt refueling operations as deemed necessary, Ensures that refuel floor activities are conducted in a professional manner.

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l Approves performance of each step during core attemations in accordance with j the Nuclear Component Transfer List (NCTL). Control of refueling activities which have the potential for aifecting core reactivity by maintaining continuous communication with the refueling bridg s QUESTION: 081 (1.00)

Which one of the following would qualify as a " Temporary Alteration" as defined in QAP 300-12,

" Temporary Alterations"? A circuit card is pulled to disable an annunciator, A hose is installed to drain a heat exchanger under an OO Installation of an electrical jumper for testing under an approved work procedure which is to be completed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, An electrical lead is lifted in accordance with a surveillance procedure which is to be completed by the end of shif i i

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l QUESTION: 082 (1.00)

Approval of a temoorary procedure is required. The Station Manager is off-site at Downers Grove and cannot return today. His signature is required for final approval of the procedure.

The task must be performed within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The Station Manager can be reached by I

facsimile and by telephone Which of the following is correct? The procedure can be signed by anyone designated by the Station Manager.

i The procedure will have to wait until the Station Manager retums to Quad Cities Statio Fax the procedure to the Station Manager. After the procedure has been signed, have the signed procedure faxed back to the statio Telephone the Station Manager, gain verbal approval over the telephon Document the approval on the Verbal Approval Documentation For QUESTION: 083 (1.00)

l Per QOA 0010-05, the NSO at the 2201-5(6) rack is directed to cooldown the reactor by manually closing the relay contacts of the rack. The NSO is cautioned not to actuate any single relief valve within 10 seconds after that valve has been close The 10 second delay between relief valve openings is used to prevent excessive containment loads due to: steam in the relief valve discharge lin a low water level in the relief valve discharge lin a high water level in the relief valve discharge lin a two phase mixture in the relief valve discharge line.

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SENIOR REACTOR OPERATOR Page 45 QUESTION: 084 (1.00)

An inadvertent Group One isolation occurs causing reactor pressure to peak at 1355 psig.

Select the required action (s), if an Notify the NRC within four hour Notify the Site Vice President within one hour, Do NOT restart the unit until authorized by the NR Hot shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and notify the NRC within four hours.

QUESTION: 085 (1.00)

PRIMARY CONTAINMENT CONTROL, QGA 200, has an override step that directs Drywell sprays to be stopped if pressure drops BELOW 2.5 psig. Which ONE of the following statements describes the reason for this step? .5 psig corresponds to 180*F, so there is no need to continue Drywell spray This action ensures drywell to wetwell design differential pressure is not exceede This ensures the Drywell structure doesn't endure excessive thermal stresses due to rapid cooldow This is done to maintain a positive torus pressure to prevent air from being drawn in through the reactor building to torus vacuum breake I

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SENIOR REACTOR OPERATOR Page 46 QUESTlON: 086 (1.00)

In order to perform an RPV blowdown using the ADS valves, QGA 500-1, RPV BLOWDOWN, requires torus level above 5 feet. WHICH ONE of the following is the reason for this restriction? To ensure sufficient water in the torus to absorb tha heat from the reactor, To ensure the ADS valve discharge T-quenchors are submerged to prevent direct pressurization of the torus.

l l To ensure sufficient water in the torus for subsequent reactor water level restoration using core spray and LPC To ensure sufficient water in the bottom of the torus to prevent the steam from the relief valves from impinging on the bottom of the toru QUESTION: 087 (1.00)

Which ONE of the following sets of plant parameters will cause the Main Turbine to trip? EHC Control Oil Pressure 1300 psig Turbine Speed 1930 rpm Condenser Vacuum 23 in Hg Bearing Oil Header Pressure 15 psig EHC Control Oil Pressure 1200 psig Turbine Speed 1925 rpm Condenser Vacuum 25 in Hg Bearing Oil Header Pressure 12 psig EHC Control Oil Pressure 1200 psig Turbine Speed 2008 rpm Condenser Vacuum 24 in Hg Bearing Oil Header Pressure 13 psig EHC Control Oil Pressure 1500 psig Turbine Speed 1920 rpm Condenser Vacuum 22 in Hg Bearing Oil Header Pressure 10 psig

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SENIOR REACTOR OPERATOR Page 47 QUESTION: 088 (1.00)

Following a reactor scram the operator is directed to stabilize reactor pressure below 1060 psig using the main turbine bypass valves. IDENTIFY the reason for this upper pressure limi Maintaining pressure below this point: minimizes positive reactivity additions, minimizes RPV level control problem allows the scram logic to be reset when require provides positive operator control of the cooldown rat QUESTION:089 (1.00)

The Hot Shutdown Boron Weight (HSBW) is the amount of boron that will maintain the reactor shutdown under hot standby conditions with the following assumptions:

-

All rods are withdrawn to the maximum rod block limit

-

No Xenon in the core

- No voids in the core

-

Reactor water temperature at saturation temperature for lowest ADS valve lifting setpoint The HSBW assumes reactor water level to be at: /3 core heigh The high level trip setpoin I The Top of Active Fuel (TAF). j The low water level scram setpoint.

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SENIOR REACTOR OPERATOR Page 48 QUESTION: 090 (1.00)

in accordance with QGA 200-5 HYDROGEN CONTROL, if Drywell Hydrogen Concentration is  ;

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above 6% and Drywell Oxygen Concentration is above 5%, the operator is directed to vent the drywell or torus depending on torus level. Which of the following is the correct? Vent the torus if torus level is above 30 ft to maximize the effect of scrubbin Vent the Dr ywell and torus if torus level is below 30 ft to minimize the vent tim Vent the torus if torus level is below 30 ft to minimize the amount of radioactivity released.

i Vent the Drywell if torus level is below 30 ft because the torus vent lines are NOT 4 designed to accommodate the flow of wate ]

i QUESTION: 091 (1.00)

Which ONE of the following describes the effect a reactor vessel pressure signal of 1250 psig will have on the reactor recirculation pumps and alternate rod insertion (ARl) system?

The recirc pump...

i drive motor breaker will trip and the ARI solenoid valves will energize.

! drive motor breaker will trip and the ARI solenoid valves will deenergiz generator field breaker will trip and the ARI solenoid valves will energiz generator field breaker will trip and the ARI solenoid valves will de-energi_e.

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SENIOR REACTOR OPERATOR Page 49 QUESTION: 092 (1.00)

A scram signal has been processe The following plant conditions exist:

-

The SDV is full-

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The scram pilot valve air header is "0" psig

. 17 control rods are NOT fully inserted

-

Reactor power is stable at 12%

"hich ONE of the following methods should be used to insert the saining control rods? Vent the scram air heade De-energize the scram solenoid Reset the scram to drain the SDV and rescram.- Individually vent the hydraulic control unit QUESTION: 093 (1.00)

Given a cold water addition to the reactor with the unit at 100%

power, which of the following is correct concerning a power 1 reduction?

' Reduc 6'recirc flow and then insert control rods because just inserting control rods may ~ result in Linear Heat Generation Rate (LHGR) violations, Insert control rods and then reduce recirc flow because just inserting control rods may result in Minimum Critical Power Ratio (MCPR) violations.

Insert control rods and then reduce recire flow to l

maintain the difference between the APRM flow bias slope and the 100% flow control line (FCL) slop Reduce recirc flow and then insert control rods because just inserting control rods may result in Average Planar Linear Heat Generation (APLHGR) violations and possible fuel failures.

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SENIOR REACTOR OPERATOR Page 50

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- QUESTION: 094 (1.00)

QGA 200 PRIMARY CONTAINMENT CONTROL defines the Primary Containment Pressure -

Limit as the maximum primary containment pressure...

' which can occur without steam in the torus airspace.

l~ r.

L at which vent valves can be opened and closed to vent the RP which can be maintained without exceeding the torus heat removal capability load if ADS valves ope '

l l at which initiation of RPV depressurization will NOT result in exceeding the l containment design limit.

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l QUESTION: 095 (1.00)

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! Drywell flooding is in pro 0ress in accordance with QGA 500-3 DRYWELL FLOODING. WHICH )

l ONE of the following is equivalent to the Top of Active Fuel?

l l feet

! feet feet

! feet i

QUESTION: 096 (1.00)

j- - A transient on Unit 1 has resulted in reactor pressure slowly increasing. WHICH ONE of the

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following operator actions should be taken to control reactor pressure in accordance with QCOA 201-3, " Reactor High Pressure"? R** ice pressure setpoint selector.

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L Reduce the EHC Load Limit setpoin Reduce the Max Combine Flow Limit Set Potentiomete Open the main turbine bypass valves using the bypass valve opening jac l SENIOR REACTOR OPERATOR Page 51

QUESTION: 097 (1.00)

QGA 300 SECONDARY CONTAINMENT CONTROL directs the installation of jumpers to bypass Reactor Building Ventilation isolation. Which of the following will cause Reactor Building i Ventilation to isolate with the jumpers installed? I High radiation signal ONLY High drywell pressure ONLY I High drywell pressure and high radiation signal High drywell pressure and low reactor vessel level l

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l QUESTION: 098 (1.00)

Which of the following states the reason (s) for running Turbine Building ventilation when l executing QGA 400 RADIOACTIVITY RELEASE CONTROL 7 I To allow... , personnel access and to assure an elevated monitored releas personnel access and to assure a ground level monitored releas operation of Turbine Building equipment withe it exceeding max safe condition operatic.n of Turbine Building equipment without exceeding max normal condition >

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SENIOR REACTOR OPERATOR Page 52 l

l QUESTION: 099 (1.00)

Which ONE of the following is the MOST preferred indicator for Torus Temperature? Process computer point C169(C269) OR C170(C270). Tl 1(2)-1640-9, TORUS H2O TEMP point A/B on the 901(2)-4 pane Tl 1(2)-1640-200, TORUS H2O TEMPERATURE point A/B on the 901(2)-36 pane Average of points A and B on Tl 1(2)-1602-8, TORUS H2O TEMP on the 901(2)-21 pane QUESTION: 100 (1.00)

QGA 500-3, "Drywell Flooding" directs venting of the RPV when the Primary Containment water level reaches 28 fee feet is the I elevation point that corresponds to -142" reactor vessel leve elevation point that corresponde to -184" reactor vessel level, elevation of tiie lowest Recirculation System pipin j highest possible elevation at which coolant could discharge from a leak in a primary syste I ("*"*"" END OF EXAMINATION *"***"**)

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SENIOR REACTOR OPERATOR Page 53 i REFERENCE DATA ANSWER: 001 (1.00) ANSWER: 006 (1.00) ,

' REFERENCE: REFERENCE:

l Modification of QC questions 10019 and Modification of QC question 970 from

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10359 from system 7500. Modification date system 1603. Modification date 012898.

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01129 S/R/A-1603-EK012 S/R-7500-EK023 223002A302 ..(KA's)

l 261000A207 ..(KA's)  ;

l ANSWER: 007 (1.00)  ;

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ANSWER: 002 (1.00) REFERENCE:

l REFERENCE: T.S. 3. New question 011398 Objectives S/R-1100-EK031 i l S/R/A-6600-EK020 211000G005 ..(KA's) j 264000K107 ..(KA's)  !

l ANSWER: 008 (1.00)

ANSWER: 003 (1.00) i l

REFERENCE

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l REFERENCE: New question 012898 New question 012698 S/R-0702-EK021 ,

S/R-1000-EK011 215003A403 ..(KA's)

205000A205 ..(KA's)

ANSWER: 009 (1.00)  :

ANSWER: 004 (1.00) c.

l REFERENCE:

l REFERENCE: New question 012898 226001K506 ..(KA's) S/R-0263-EK022 216000A203 ..(KA's)

i ANSWER: 005 (1.00)

l ANSWER: 010 (1.00)

REFERENCE: QC exam bank, question 4278 from system REFERENCE:

160 New question S/R-1601-EK006 S/R-1000-EK013 223001A209 ..(KA's) 203000K117 ..(KA's)

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SEMOR REACTOR OPERATOR Page 54'

ANSWER: ' 011 (1.00) ANSWER: 016 (1.00)- ' REFERENCE: REFERENCE:

LIC-0203 page 38 L-QGA504 218000K102 .(KA's) Objective S/R-0001-EK055 295031G012 ..(KA's)

..

ANSWER: 012 (1.00)

a.' ANSWER: 017 (1.00)

REFERENCE: b New question REFERENCE:

QCOA 1000-02 LOSS OF SHUTDOWN S/R-0302-EK014 COOLING 201001K205 ..(KA's)

295021K102 ..(KA's)

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ANSWER: '018 (1.00)

ANSWER: 013 (1.00) ' REFERENCE:

REFERENCE: Modification of QC question 1044 system 0250. Modification date 010198 QC exam bank, question 681 from system S/R-0250-EK019 0001 239001K201 ..(KA's)

S/R-0001-EK027 2. .(KA's)

~ ANSWER: 019 (1.00) ANSWER: 014 (1.00) REFERENCE: Modification of QC questions 1574 and REFERENCE: 1166 from system 5652. Modification date QC exam bank, question 9852 from system 01079 S/R-5652-EK021 S/R-0001-EK028 241000A116 ..(KA's)

295032K104 ..(KA's)

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ANSWER: 020 (1.00) j ANSWER: 015 (1.00) REFERENCE:

REFERENCE: New Question I ( QC exam bank. question 674 system 000 S/R-2300-EK013 S-0001-EK012 QCOP 2300-09 i 295028K203 ..(KA's) 206000A104 ..(KA's) j

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SENIOR REACTOR OPERATOR Page 55 ANSWER: 021 (1.00) ANSWER: 026 (1.00) REFERENCE: REFERENCE:

QCGP 1-1, rev 19 pages 59 and 60 Modification of question for Brunswick exam 245000A302 ..(KA's) bank. Modification date 01089 S/R-05000-EK007 212000A216 ..(KA's)

ANSWER: 022 (1.00) REFERENCE: ANSWER: 027 (1.00)

Modification of QC question 884 from system 0703. Modification date 01179 REFERENCE:

. S/R-0703-EK009 New question 010898 215005K102 ..(KA's) S/R-0250-EK024 -

239001K127- ..(KA's)

ANSWER: 023 (1.00) ANSWER: 028 (1.00)

REFERENCE: New question 010898 REFERENCE:

. S/R-0702-EK016 QC exam bank 215003A106 ..(KA's) S/R-0705-EK026 215002K604 ..(KA's)

ANSWER: 024 (1.00) ANSWER: 029 (1.00)

REFERENCE: QC exam bank REFERENCE:

S/R-0701-EK022 Modification of QC question 759 from

'215004K302 ..(KA's) system 0207. Modification date 01089 S/R-0207-EK015

, 201006K101 ..(KA's)

ANSWER: 025 (1.00) REFERENCE: ANSWER: 030 (1.00)

New Question 010898 QCOP 1700-05 rev 4 page 2 REFERENCE:

< 272000K103 ..(KA's) QC exam bank question 1009 from system 0210 S/R-0201-EK022 290002A201 ..(KA's)

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SENIOR REACTOR OPERATOR Page 56 i ANSWER: 031 (1.00) ANSWER: 036 (1.00) I REFERENCE: ' REFERENCE:

QC exam bank. question 1971 from system QC exam bank, modified the distractors 0001 020698 -

S/R-0001-EK061 QCAP 0630-06 rev 5 page 1 K102 ..(KA's) 2. ..(KA's)

ANSWER: 032 (1.00) ANSWER: 037 (1.00) REFERENCE: REFERENCE:

QC exam bank QC exam bank. question 30 from system QAP 0300-28 rev 15 page 1 190 . ..(KA's) S/A-1900-EK021 QCOP 6600-11 233000G009 ..(KA's)

ANSWER: 033 (1.00) REFERENCE: ANSWER: 038 (1.00)

QC exam bank QAP 300-13 rev 24page 2 REFERENCE:

2.2.13' ..(KA's) QC exam bank. question 25 from system 5751

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S/R-5751-EK021 QCOP 5750-02 ANSWER: 034 (1.00) 288000K402 ..(KA's)

d.

i REFERENCE:

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QC exam bank

! QCAP 2400-03 rev 4 pages 1 & ANSWER: 039 (1.00) i 2. .(KA's) REFERENCE:

QC exam bank NO-421-TPO-02 001 ANSWER: 035 (1.00) 2.1.29 ..(KA's) REFERENCE:

QC exam bank, modified distractors i

QCAP 0890-11 ANSWER: 040 (1.00)

2. ..(KA's) REFERENCE:

QOA 0010-05, Plant Operations with the Control Room Inaccessible 295016G010 ..(KA's)

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l SENIOR REACTOR OPERATOR Page 57 ANSWER: Ni (1.00) ANSWER: 046 (1.00) REFERENCE: REFERENCE:

LIC-3700 page 22 taken from the last QC QCOA 0202-01 JET PUMP / SHROUD exa ACCESS COVER FAILURE 2.4.24 ..(KA's) 295001K102 ..(KA's)

I ANSWER: 042 (1.00) ANSWER: 047 (1.00) REFERENCE: REFERENCE:

Modification of question from BNP exam QC exam bank question 1951 from system bank. 020498 4700 I OOA 3300-02 LOSS OF CONDENSER S/R-0250-EK023 VACUUM 295019K205 ..(KA's)

295002A201. ..(KA's) ,

I ANSWER: 048 (1.00)

ANSWER: 043 (1.00) REFERENCE:

REFERENCE: New question 020598 New question 020498 S/R-0001-EK022 l QOA 6800-03 120/240 VAC ESSENTIAL 295013K302 ..(KA's)

SERVICE BUS FAILURE 295003K305 ..(KA's)

ANSWER: 049 (1.00)

b, ANSWER: 044 (1.00) REFERENCE: New question REFERENCE: QCOA 0300-01 CONTROL ROD DRIVE New question PUMP FAILURE

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QOA 6900-04 295022K301 ..(KA's)

295004A102 ..(KA's)

i ANSWER: 050 (1.00)

ANSVER: 045 (1.00) c.

i a REFERENCE:

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REFERENCE: modification of QC question 987 from T.S. Base system 1603.

I S/R-2300-EK009 A-1603-EK011 295008K101 ..(KA's) 295020K204 ..(KA's)

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SENIOR REACTOR OPERATOR Page 58 ANSWER: 051 (1.00) ANSWER: 056 (1.00) d.

REFERENCE: REFERENCE:

New question QC exam bank S/R-0001-EK033 QEP 9900-102 295017K303 ..(KA's) 2.1.19 ..(KA's)

ANSWER: 052 (1.00) ANSWER: 057 (1.00) a.

REFERENCE: REFERENCE:

New question QC Exam bank QCAP 0630-06 rev 5 page 5 QCAP 044-016 l 2. .(KA's) 2.1.30 ..(KA's)

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ANSWER: 053 (1.00) ANSWER: 058 (1.00) b.

REFERENCE: REFERENCE:

New question New question QAP 0300-02 rev 51 page 19 QCAP 0200-10 EMERGENCY 2.1.30 ..(KA's) OPERATING PROCEDURE EXECUTION STANDARDS rev 16 pages 4,5,6& . ..(KA's)

ANSWER: 054 (1.00) l REFERENCE: ANSWER: 059 (1.00)

New question b.

QCAP 0230-04 rev 20 page 2 & 3. REFERENCE:

2.2.13 ..(KA's) QC exam bank S/R-0001-EK0333 State the enty l conditions to QGA 400 Rad Release Control.

ANSWER: 055 (1.00) 295024K101 ..(KA's)

c.

REFERENCE: ANSWER: 060 (1.00))

QC exam bank c.

QCAP 1100-12 REFERENCE:

2.1.23 ..(KA's) QC exam bank 2.4.47 ..(KA's)

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' ANSWER: 061 (1.00) ANSWER: 066 (1.00) REFERENCE: REFERENCE:

New question 022598 QC exam bank. Question 17 from system S/R-0001-EK061 Given QGA 101 RPV 130 Control (ATWS) Explain the reason for the S/R-1300-EK029 actions 2.1.12 ..(KA's)

S/R-0001-EK005 Introduction to QGAs, describe the procedure use guideline K103 ..(KA's)

ANSWER: 067 (1.00) '

REFERENCE:

ANSWER: 062 (1.00) QC exam bank, question 5 from system .

REFERENCE: S/R-0202-EK021  ;

New question 022598 202001K305 ..(KA's)

S/R-0001-EK023 l 2.4.14 ..(KA's)

ANSWER: 068 (1.00) ANSWER: 063 (1.00) REFERENCE: QC exam bank. Question 40 from system REFERENCE: 020 I S/R-0001-EK061 S/R-0202-EK035 295030G004 ..(KA's) 202002K103 ..(KA's)

i ANSWER: 064 (1.00) ANSWER: 069 (1.00) > REFERENCE: REFERENCE:

QOA 0010-05 QC exam bank. Question 131 from system Objective S/R/A/B-EVAC-PK009 0202 295016K202 ..(KA's) S/R-0202-TP004 QCOA-0202-4 page 6,7 202002K302 ..(KA's)

ANSWER: 065 (1.00)

' ANSWER: 070 (1.00)

REFERENCE: QC exam bank. Modified distractor REFERENCE:

Question 32 from system 140 QC wam bank. Question 17 from system S-1400-EK027 6500 T.S. S-6500-EK032 2.1.12 ..(KA's) 226001K201 ..(KA's) l l

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SENIOR REACTOR OPERATOR Page 60 ANSWER: 071 (1.00) ANSWER: 076 (1.00) REFERENCE: REFERENCE:

QC exam bank, Question 66 system 1601 QCAP 1800-01 SHUTDOWN RISK 290001K401 ..(KA's) ACTIVITIES l Objectives S/R/B/FH/LW-RISK-K003 271000A408 ..(KA's)

ANSWER: ' 072 (1.00) REFERENCE: ANSWER: 077 (1.00)

Modification fo QC question 16 from system . Changed one of the distractor REFERENCE:

S/R-5752-EK001 Modification of QC questions 4279 and 2.1.27 ..(KA's) 5854 from system 2300. Modification date 010798 S-2300-EK033 2.1.27 ..(KA's)

ANSWER:' 073 (1.00) REFERENCE:

QCOP 1200-11 page 2 ANSWER: 078 (1.00)

LIC-1200 j 204000G010 ..(KA's) REFERENC New question 010898 S/UFH-0803-EK013 ,

234000K502 ..(KA's)  !

ANSWER: 074 (1.00)  ; !

REFERENC i QC exam bank, question G from system AJSWER: 079 (1.00)

100 ,

S/R-1000-EK016a REF(RENCE: )

203000K409 ..(KA's) Modifiution of QC exam bank question i 1945 syvem 1900. Modification date 01099 S/R-1900-Er '126 ANSWER: 075 (1.00) UF-1900-EKO. S j

. (KA's)

' A103 REFERENCE:

QC exam bank, question 15 from system l 170 ANSWER

Cs0 (1.00)

S/R-1701-EK020 a.

l 271000A408 ..(KA's) REFERENCE:

New question QAP 0300-38 rev 5 pages 2 & .2.26 ..(KA's)

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! SENIOR REACTOR OPERATOR Page 61 ANSWER: 081 (1.00) ANSWER: 086 b REFERENCE: REFERENCE

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' QAP 0300-12 page 2,3 2.4.20 ..(KA's)

Temp Alts OJT/OJE 294001K102 ..(KA's) j ANSWER: 086 (1.00) i ANSWER: C s2 (1.00) REFERENCE- I QC question 1991 from system 0001 REFERENCE: S/R-0001-EK040 QC exam bank 295030K208 ..(KA's)  ;

QAP 0200-19 rev 4 page 1 1 2.2.11 ..(KA's)

ANSWER: 087 (1.00) '

l ANSWER: 083 (1.00) REFERENCE: New Question 020298 l REFERENCE: S/R-5651-EK009 QCOP 0203-01 295005A201 ..(KA's)

Objectives S/R/A/B-EVAC-PK003

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295016G006 ..(KA's)

l ANSWER: 088 (1.00) ANSWER: 084 (1.00) REFERENCE: Modification question from FITZ exam ban REFERENCE: 020298

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, T/S Section 6.7 SAFETY LIMIT S/R -0001-EK061 VIOLATIO K207 ..(KA's)

Modification of QC exam bank question.

l The stem and two distractors were modifie ANSWER: 089 (1.00) REFERENCE:

( ANSWER: 085 (1.00) Modification for BNP exam bank. 020298 S/R-0001-EK058 REFERENCE 295037K104 ..(KA's)

Modification of QC quesiton 678 system 0001. Modification date 013198 S/R-0001-EK022 l

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W -v SENIOR REACTOR OPERATOR Page 62 ANSWER: 090 (1.00) ANSWER: 095 (1.00) REFERENCE: REFERENCE:

New question 020298 New question 020298 S/R-0001-EK020 S/R-0001-EK050 500000K208 ..(KA's) 2.4.24 ..(KA's)

.

ANSWER: 091 (1.00) . ANSWER: 096 (1.00)

i 4 d.

l REFERENCE: REFERENCE:

l New question Modification of a question used on the last S/R-0202-EK009 QC Initial Licensing Exam. Distractors a 295025A107 ..(KA's) and c were changed.

l QCOA 201-3, Reactor High Pressure, Rev.

l 6, Subsequent Operator Action C.3.c (page

'

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ANSWER: 092 (1.00) 295007A105 ..(KA's)

, c.

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REFERENCE:

QCOP-28 ALTERNATE CONTROL ROD INSERTION ANSWER: 097 (1.00) .

REFERENCE: )

New question 020398 J ANSWER: 093 (1.00) S/R-0001-EK029 a 295033K201 ..(KA's)

REFERENCE:

New question 020398 QCOA 0400-01 REACTIVITY ADDITION

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rev 7 page ANSWER: 098 (1.00)

I 295014A107 ..(KA's) REFERENCE:

Modification of QC question 1994 from system 0001, 020498 l ANSWER: 094 (1.00) S/R-0001-EK035 l A106 ..(KA's)

i REFERENCE:

! New question 020398 S/R-0001-EK022 2.4.17 ..(KA's)

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SENIOR REACTOR OPERATOR Page 63 ANSWER: 099 (1.00) REFERENCE:

QC exam bank. question 5524 from system 0001.

I S/R-0001-EK007 295026A103 4.(KA's)

ANSWER: 100 (1.00) REFERENCE:

Modification of QC question 5550 from system 0001. Modification date 020498 S/9.-0001-EK0505 Given QGA 500-3 tExplain the reasons for the action K207 ..(KA's)

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(""""" END OF EXAMINATION "*"*"")

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SENIOR REACTOR OPERATOR Page 64 ANSWER KEY l

001 a 021 b 041 a 061 b 081 a 002 a 022 c 042 b 062 d 082 c l 003 b 023 a 043 a 063 b 083 c 004 b 024 b 044 a 064 c 084 c

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005 c 025 d 045 a 065 d 085 b 006 a 026 b 046 c 066 c 086 b 007 a 027 c 047 a 067 a 087 c 008 a 028 c 048 b 068 d 088 c

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009 c 029 b 049 b 069 c 089 b 010 d 030 d 050 c 070 a 090 c 011 a 031 b 051 b 071 a 091 c 012 a 032 c 052 c 072 d 092 c i 013 a 033 d 053 d 073 a 093 a 014 d 034 d 054 d 074 c 094 b I

015 d 035 b 055 s h 075 b 095 a b

016 a 036 a 056 d 076 c 096 d s 017 b 037 a 057 a 077 b 097 a 018 a 038 a 058 b 078 c 098 a 019 c 039 d 059 b 079 c 099 b 020 a 040 d 060 c 080 a 100 c l

l (""""" END OF EXAMINATION *"*"*"*)

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l LIMITED SENIOR REACTOR OPERATOR Page 2 of 30 ANSWER SHEET

' Multiple Choice (Circle your choice) I t if you change your answer, write your selection in the blan MULTIPLE CHolCE

' 001 a . b c d 023 a b c d 002 a b c d 024 abcd 003 a b c d 025 a b c d

004 a b c d 026 a b c d 005 a b c d 027 a b c d 006 a b c d 028 a b c d

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l 007 a b c d 029 a b c d 008 a b c d 030 a b c d 009 a b c d 031 abcd 010 a b c d 032 a b c d

011 abcd 033 a b c d 012 a b c d 034 abcd 013 a b c d 035 a b c d 014 a b c d 036 a b c d 015 a b c d 037 a b c d j 016 a b c d 038 a b c d 017 a b c d 039 a b c d 018 a b c d 040 a b c d 019 a b c d 041 abcd 020 a b c d 042 a b c d 021 abcd 043 a b c d 022 a b c d 044 abcd l \

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c l LIMITED SENIOR REACTOR OPERATOR Page 3 of 30 l

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i ANSWER SHEET Multiple Choice (Circle or X your choice)

1 If you change your answer, write your selection in the blan MULTIPLE CHOICE 045 a b c d 046 a b c d 047 a b c d 048 a b c d 049 a b c d 050 a b c d l

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(.........* END OF EXAMINATION *"*****")

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l LIMITED SENIOR REACTOR OPERATOR Page 4 of 30 (Read VerbatimJ 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 80.00 percent or greater. Every l question is worth one point.

i The time limit for completing the examination is four hour . You may bring concils and calculators into the examinatic.. com. Use only pencils to ensure correct scan-tron gradin . Your name is printed in the blank provided on the examination cover sheet. Print your name on the scan-tron answer sheet.

l Mark your answers on the answer sheet provided. If you decide to change your original l answer, erase completely and blacken the correct answer If the intent of a question is unclear, ask questions of the NRC examiner or the designated facility instructor onl . Restroom trips are permitted, but only one applicant at a time will be allowed to leave.

l Avoid all contact with anyone outside the examination room to eliminate even the

! appearance or possibility of cheating.

I l When you complete the examination, retum the coversheet and scantron answer sheet to the NRC examiner. Leave everything else at your table. Remember to sign the statement on the examination cover sheet indicating that the work is your own and that

you have neither given nor received assistance in completing the examination. Any

, scrap paper will be disposed of immediately after the examination.

!

l 1 After you have turned in your examination, leave the examination area as defined by the proctor or NRC examiner. If you are found in this area while the examination is still in progress, your license may be denied or revoked.

l 1 Do you have any questions?

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l LIMITED SENIOR REACTOR OPERATOR Pags Gvi30 l

QUESTION: 001 (1.00)

The following indication are available:

-

Reactor cavity level is flooded and maintaining a steady leve The dryer / separator pit water is decreasing slowl The dryer / separator pit sump pump is not runnin The dryer / separator pit liner leak flow sightglass indicates flo No ARMS are alarming These are indications of: the dryer / separator pit is leaking into the drywell via the refueling bulkhea the reactor bellows are leaking, draining the dryer / separator pit into the reactor cavity, the dryer / separator pit has developed a leak and is leaking to the reactor building floor drain sum i the reactor bellows has developed a leak and is leaking to the reactor building equipment drain tank.'

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i LIMITED SENIOR REACTOR OPERATOR - Page 6 of 30 l

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- QUESTION: 002. (1.00)

As part of a maintenance job being performed on the reactor head, you notice two mechanical maintenance department personnel relocating six lead blankets to better reach their task. What action should you take? Stop the job, only station laborers are allowed to move portable shieldin Stop the job as Rad Protection approval must be obtained prior to relocating portable shieldin Contact the cognizant maintentnce supervisor to determine if this is part of their work package, Allow the Job to continue, individual workers can relocate shielding as long as they do a radiation surve QUESTION: 003 (1.00)

During refueling operations on the refuel floor you notice that there is a high level in the fuel pools. It appears that the fuel pool couling pumps have tripped, (there is no flow across the skimmer weirs). What is a possible outcome if water level continues to increase? The excess fuel pool water will leak into the on-line reactor's reactor cavity and flood the drywel The fuel pool ventilation ports may flood out allowing fuel pool water to enter the l reactor building ventilation ductwor ! The fuel pool skimmer surge tanks may overflow causing a contamination hazard l on the third floor of the reactor buildin I There is no problem for fuel handling activities since this higher level provides -

additional shielding to any fuel assemblies being transferre i

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l LIMITED SENIOR REACTOR OPERATOR Page 7 of 30 QUESTION: 004 (1.00)

You are supervising reactor cavity decontamination after refueling and draindown. A final walkthrough is needed. You observe that your digital dosimeter indicates that you have received 70 MREM accumulated dose today. How much additional exposure are you allowed to receive per Quad Cities Station Administrative guidelines? MREM TEDE MREM DDE MREM DDE MREM TEDE QUESTION: 005 (1.00)

The plant has just completed a shutdown for refueling. Vessel disassembly has commence The IM department has determined that IRM 15 and SRM 21 are inoperable. The detectors have ceased to function. What action (s) must be completed prior to full core offload? Shutdown margin must be demonstrate SRM 21 will have to be replaced so offload can occur in that quadran l Both instruments must be replaced before any core alterations can begi ! IRM 15 must be restored in order to meet the minimum operable channel requirement :

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LIMITED SENIOR REACTOR OPERATOR Page 8 of 30 QUESTION: 006 (1.00)

The following conditions exist:

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Fuel offload is in progres All SRM/lRM/APRM channels are operabl One control rod is stuck at position 0 Two SRM ' Shorting Links" are removed in accordance with QCFHP 0100-01 for core unloa No fuel loading chambers (Dunkers) are installe All quadrants of the core still have fue Which one of the following describes when the remaining two " Shorting Links" must be removed in order to continue core alterations? If average SRM counts dout If one SRM becomes inoperabl Immediately; prior to continuing with core alteration When two IRMs in the same trip system become inoperabl QUESTlON: 007 (1.00)

Which one of the following is considered a heavy load lift that would require a special procedure and approval prior to the lift taking place? Movement of any loads with the 125 ton hoist block.

f Handling of any load greater than 2000 lbs with the 9 ton boo Handling of any loads with the 9 tcn book over restricted area Movement of any loads greater than 250 lbs with the jib crane over restricted areas.

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LIMITED SENIOR REACTOR OPERATOR Page 9 of 30 l )

l QUESTION: 008 (1.00)

QCFHP 0400-06 Precaution D.1 requires that upon completion of uncoupling a CRD from it's blade that the associated HCU is valved out. Why is this done? To prepare the CRD drive unit for replacemen To ensure the rod operability Tech Spec is satisfie To prevent equipment damage in the event the rod is scramme To prevent inadvertent draining of the reactor vessel if the CRD drive unit is remove QUESTION: 009 (1.00)

While performing refueling interlock checks, the 1000 pound test weight was accidentally l dropped into the fuel pool. The weight damaged both new fuel and spent irradiated fuel as it

- travelled to the bottom of the pool. A cloud of bubbles rose to the surface of the pool. The ,

refuel floor ARMS did not alarm. Which of the following is the most correct course of action? ! The damage to the new fuel is the primary radiological concem. The refuel floor should be evacuated and surveys done to prevent the spread of contaminatio '

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' If the refuel floor ARM's alarm than evacuate the refuel floor, in the mean time, stop all fuel movements and notify the Fuel Handling Supervisor, Shift Engineer, NSO, and Qualified Nuclear Engineer,  ;

l The refuel floor ARMS will not respond to Kr-85 noble gas which is released from irradiated fuel. Stop all fuel movement, evacuate the refuel floor, notify the Shift Engineer and request that he evacuate the reactor building as wel ,

i i The damage to the irradiated fuel is the primary radiological concem. If the refuel floor ARM's alarm then evacuate both the refuel floor and the reactor building. In the mean time, notify the Fuel Handling Supervisor, Shift Engineer, NSO, and Qualified Nuclear Engineer.

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LIMITED SENIOR REACTOR OPERATOR Page 10 of 30 QUESTION: 010 (1.00)

A recent change to QCFHP 0400-06 has added a step to vacuum each CRD guide tube whenever the blade is removed. Why was this step added? The step was adopted in an effort to lower the overall radiation levels present in the vesse The step is intended to assist in lowering the radiation levels found in the inlet CRD strainer Some plants that had shroud repairs done by General Electric have been experiencing slow CRD scram times due to swarf cloggin Some plants report that vacuuming enhances the guide tube seating of CRD blades thus reducing under vessel leakage during maintenanc QUESTION: 011 (1.00)

You are attempting to insert a fuel assembly into a cell containing three other previously loaded Lel assemblies. The assembly you are loading binds in place shortly before it seats fully in the core. Which of the following conditions would be suspected and checked to determine the source of the binding condition? Assembly lower tie plate caught on upper gri Assembly lower tie plate caught on control ro Spacer buttons in direct contact with each othe Channel fastener in direct contact with the upper gri :

l l LIMITED SENIOR REACTOR OPERATOR Page 11 of 30 OliESTION: 012 (1.00)

While lifting a spent fuel cask from the reactor building ground floor level to the cask washdown area on the 690 ft level, it is observed by one of the fuel handlers that the bottom of the fuel cask is greater than one foot above the refuel floor. What action should be taken? The cask must be grounded immediately and the failed limit switch repaire Verify that the restricted mode limit switches are set properly and continu The up-limit switch needs to be adjusted higher so it will not interfere with cask movement operation Operation can continue for up to forty eight hours provided that an operator is on the floor to ensure the cask is within the restricted zon QUESTION: 013 (1.00)

The Overhead Crane functions have been automatically disabled (shutdown) by the equalizer circuit. It has been determined that the length of the redundant cable trains is unequa What Overhead Crane functions are enabled (restored) if the EQUAUZER BYPASS SWITCH is positioned to BYPASS 7 All crane functions... are restore EXCEPT raise and lower are restore EXCEPT trolley left and right are restored.

l EXCEPT bridge forward and reverse are restored.

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LIMITED SENIOR REACTOR OPERATOR Page 12 of 30 QUESTION: 014 (1.00)

Fuel movements are in progress. The Reactor mode switch is locked in the REFUEL position and all control rods are fully inserted into the reactor core. A fuel assembly has been grappled l

in the fuel pool and is being transferred to the reactor core for installatio Which statement correctly describes the refueling interlock (s) and/or light indication (s) that will -

be lit on the refuel bridge when the refueling bridge is positioned near or over the reacto vessel? Rod block #1 ONLY Rod block #1 and #2 ONLY Bridge reverse motion stop #1 Rod block #1 and bridge reverse motion stop # QUESTION: 015 (1.00)  ;

'1 The core assembly components are located by the use of an X-Y coordinate system. Unit 1 and Unit 2 reference poir.t for the start of this coordinate system (position 00-00)'is which corner of the reactor vessel? North west ' i South west North east l South east

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, I LIMITED SENIOR REACTOR OPERATOR Page 13 of 30 QUESTION: 016 (1.00)

The Unit-2 refueling platform is over the core withdrawing the first fuel assembly of the offload.

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it is halfway out of the core. The NSO observes a single control rod withdrawn to position 04.

l No interlocks are active. What might you suspect to be the problem? Rod select power is de-energize The reactor mode switch is in shutdown.

l The refueling interlock test switch in Ge 902-28 panel is depresse There has been a failure of the full-in limit switch on the subject control ro QUESTlON: 017 (1.00)

What is the correct action with a fuel assembly half way into the core when a criticality excursion is observed in the control room? Inject Standby Liquid Control flui Immediately stop all fuel movement . Remove the fuel assembly and retum it to the fuel poo J l Complete the fuel move you are on and report to your assembly are l

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LIMITED SENIOR REACTOR OPERATOR Page 14 of 30 QUESTION: 018 (1.00)

If difficulty is encountered while installing the channel fastener due to the upper tie plate being tilted, what should be done?

a Push down slowly on the fuel assembly by the upper tie plate bail. This should return the upper tie plate to its original positio;;

I Lift slightly on the fuel assembly by the upper tie plate bail. This should return the i upper tie plate to ites original positio . Push slowly downward on the channel until the upper tie plate channel fastener post is just contacted by the channel clop. This should return the upper tie plate to its original positio Lift the channel until all pressure is removed from the upper tie plate and then push slightly on the fuel assembly by the upper tie plate bail. This should retum the upper tie plate to its original positio QUESTION: 019 (1.00)

With the mode switch in STARTUP, which of the following conditions will provide a control rod

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block? Fuel grapple hoist loade Bridge reverse motion stop #1 ' Safety travelinterlock activate Refueling platform near or over cor F LIMITED SENIOR REACTOR OPERATOR Page 15 of 30 QUESTION: 020 (1.00)

During a refueling outage it is noted that the following conditions exist:

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X co-ordinate indicates +41

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Y co-ordinate indicates +22

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Z co-ordinate indicates -003

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The mode switch is in REFUEL Which of the following conditions will exist? Rod Block #1 Rod Block #2 Reverse motion stop #1 Reverse motion stop #2 QUESTION: 021 (1.00)

What is the primary function of the monorail hoist jamming button? To assist in keeping the cable wrapped correctly on the hoist drum, To serve as an operator aid in stopping the hoist before full up is reache To ensure that highly radioactive control rod blades are not lifted near the surface of the wate To ensure that tools used for in-core maintenance do not hang up under the upper grid when extracted from the cor i I

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LIMITED SENIOR REACTOR OPERATOR Page 16 of 30 QUESTION: 022 (1.00)

If one of the restricted mode limit switches on the east side of the pathway parallel to the unit-1 fuel pool were to fail, what could be the result? Fuel cask movement would sto The spent fuel cask could traverse over the spent fuel, The bridge and trolley travel limit bypass function will still allow motio Once outside of the restricted mode pathway the hoist's downward motion is blocke QUESTION: 023 (1.00)

When moving a fuel loading chamber (Dunker) during core reload, which one of the following !

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movements of the dunker may be performed without specifically being called for on the Nuclear Component Transfer List?

< Lifting the chamber for the SRM operability surveillanc Moving the chamber within the % to 3/4 core height regio Adjusting the chamber height up or down by loss than one foo Moving the chamber at a fixed level within the same quadrant of the core.

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l LIMITED SENIOR REACTOR OPERATOR Page 17 of 30 QUESTION: 024 (1.00)

During refueling operations on Unit-2 the following conditions are noted:

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A fuel assembly is halfway loaded in the core and going dow Rod select power on the 902-5 panel is los What will be the effect on refueling operations? Rod block #1 occur Hoist motion is blocke The fuel move can be complete Refueling platform reverse motion stop #1 occur QUESTION: 025 (1.00)

An event occurs causing the unit-2 reactor cavity to drain rapidly during a refueling outage. The i

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Unit-2 keyway gates and the transfer canal gates are removed. With no operator actions, at what point will level in the unit-2 fuel pool stop decreasing? The fuel pools will both drain dr At the Tech Spec minimum of 33 fee At the bottom of the fuel pool suction line to the FPCC pump At the level of the spent fuel bail handles in the spent fuel rack .

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l l LIMITED SENIOR REACTOR OPERATOR Page 18 of 30 QUESTION: 026 (1.00)

If the reactor mode switch is locked in the REFUEL position and the bridge is over the core, which of the following correctly describes the conditions that will cause the FUEL HOIST INTERLOCK light on the Interlock Status Display panel to be lit? Any hoist is loaded and all control rods are inserte the main hoist is loaded and a control rod is withdraw the main is boist loaded and all control rods are inserte A controlis rod withdrawn and the main and auxiliary hoists are unloaded, i

QUESTION: 027 (1.00)

Which of the following correctly states a criteria to be used when verifying proper fuel assembly orientation in the reactor? Channel fasteners are at the center of the control cell.

, Channel spacer buttons press against the outside of the cel All fuel assembly serial numbers can be read looking from the sout All bail handles line up in a north-south direction across the core to maintain core symmetry.

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LIMITED SENIOR REACTOR OPERATOR Page 19 of 30 QUESTION: 028 (1.00) j What is the primary function of the raised wall section of the dryer separator pit that divides the pit from the reactor cavity? For shielding of personnel working in the drywell are To shield personnel working on the lower levels of the reactor building from radiation hazards on the refuel floor, To maintain a water seal over both the separator and the dryer to minimize radiation airborne hazards on the refuel floo I To maiatain some water shielding over the shroud head section of the separator package t.: the event of a loss of cavity leve QUESTION: 029 (1.00)

If an air line feeding the grapple jaws separates during the movement of a blade guide out of ;

the core, what effect will this have on the completion of the move? I The loss of air pressure will have no effect on the operation of the grappl The grapple will fail open as soon as the weight of the blade guide is remove The grapple will remain closed until air pressure is restored and then it will ope The grapple will fail as-is but can be released manually after the move is complete LIMITED SENIOR REACTOR OPERATOR Page 20 of 30 t

QUESTION: 030 (1.00)

During a core reload, a Fuel Handling operator was inserting an assembly into a defueled control cell.1 The assembly bumpsct tne control rod and moved it from position 00 to position 0 What refueling interlocks Gre/Aoui.1 be in effect at this time? Fault Lockout, Heist Full Down Limit, and Rod Block # Fuel Hoist interlock, Bridge Reverse Motion Stop #1, and Rod Block # Fuel Hoist Interlock, Bridge Reverse Motion Stop #2, and Rod Block # Fuel Hoist interlock, Bridge Reverse Motion Stop #1, and Bridge Reverse Motion Stop # QUESTION: 031_(1.00)

Equipment handling in the spent fuel pools is limited to the normal weight limits of each hoist with certain exceptions. What are these movement exceptions? fuel cask reactor building overhead crane

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- high density fuel storago racks interlock test weights and pool gates.

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LIMITED SENIOR REACTOR OPERATOR Page 21 of 30 QUESTION: 032 (1.00)

The following conditions exist:

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Core off load is in progres An assembly being raised is 3/4 out of the cor l

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The assembly is discovered to be the wrong one per the Nuclear Component Transfer List (NCTL).

What actions are required? Immediately lower and unlatch the assembly, then notify the reactor operator to log the evolutio Complete the off load of the fuel assembly. Secure all core alterations and notify i the Nuclear Enginee Stop all fuel moves and notify the Unit Supervisor and obtain permission to lower and uniatch the assembl j d.- Complete the off load of the fuel assembly and update the NCTL to reflect the move. Request permission to reposition the assembly back into the vesse !

QUESTION: 033 (1.00)

A fuel assembly is being transported from the fuel pool to the reactor. The platform has just entered the keyway when it is observed that the water level is going down rapidly at about two feet per minute. What action would you direct the fuel grapple operator to take? Continue into the core and complete the fuel mov Stop the fuel move in the cattle chute and evacuate the refuel floo Exit the platform and remain by the Rad tech desk to support level recovery pfforts, Reverse the platform and place the fuel assembly in the nearest open location in the fuel pool.

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QUESTION: 034 (1.00)

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' As a Fuel Handling Supervisor performing fuel moves during a refueling outage, your work schedule is as follows:

Monday = 7:00 am - 3:00 PM Tuesday = 7;00 am - 3:00 PM Wednesday = 3:00 PM - 11:00 PM ,

Thursday = 7:00 am - 3:00 PM .

Friday = 7:00 am - 3:00 PM Saturday = 3:00 PM - 11:00 PM Sunda = 7:00 am - 11:00 PM With no shift turnover time and without violating any limits, which of the following is acceptable? You can work overtime 7:00 am - 3:00 PM on Wednesda ., You can work overtime 6:00 am - 7:00 am on Thursday mornin You can work overtime 3:00 PM -- 12:00 am (midnight) on Frida You can work overtime 3:00 PM - 11:00 PM on Monday AND 3:00 PM - 4:00 )

. PM on Thursda QUESTION: 035 (1.00)

During refueling operations with the mode switch in refuel it is determined by the IM department that the one-rod-out interlock has become inoperable. What action must be taken per the Technical Specifications? Core alterations must be suspende Core alterations can continue for seven day , Refueling operations are permissible provided the mode switch is locked in REFUE Refueling operations are permissible provided the mode switch is locked in !

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QUESTION: 036 (1.00) i What is the Tech Spec basis for the fuel pool minimum level? ensures adequate cooling capabilit Ensure sufficient radiation shieldin This is the minimum required to remove iodin i minimum required for fuel pool cooling pump suction capabilit QUESTION: 037 (1.00)

Wnich of the following is a correct statement conceming the Tech Spec bases regarding refueling design features and procedures? The two methods of safety function independently from each othe Plant refueling procedures augment the designed refueling safety feature Plant procedures serve as an enbrsncement to the original refueling interlock )

i The refueling interlocks serve as a backup to the plant refueling procedure I i

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F LIMITED SENIOR REACTOR OPERATOR Page 24 of 30 QUESTION: 038 -(1.00)

A fuel assembly is being removed from the reactor core when the following indications are observed by the grapple operator:

- Hoist Jam light illuminate Hoist motion stopped The grapple operator states that he believes the hoist is mis-aligned in relation to the fuel bundle. What action should be taken?

ii Completely reinstall the fuel bundle into the core and attempt the move again in slow speed to get pest the bindin Reposition the hoist by moving the bridge and trolley with' the jog system and then attempt to lift the assembly agai Inform the Fuel Handling Supervisor, Shift Engineer, and Nuclear Enginee Upon their consent attempt to make/ continue the fuel mov Stop all fuel movements by leaving the bundle where it is until a complete engineering analysis is made of the situation and plant management concurs with continuing fuel movemen QUESTION: 039 (1.00)

-F During core off loading operations on Unit-2 SRM 22 is determined to be inoperable by the .

control room.' SRM 21 and 23 were previously out of service awaiting detector replacemen What operational impact does this have? Core off loading may continue in the adjacent quadrant which has an operable SR All core off loading operations must be discontinued and Dunking Chambers must be used, All core off loading must be discontinued and at least one more SRM must be restore Core off loading may continue as long as the SRM in an adjacent core quadrant remains operabl I

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LIMITED SENIOR REACTOR OPERATOR Page 25 of 30 l QUESTION: 040 (1.00)

A step was recently added to QCFHP 0200-02 that restricts to 3 fuel bundies any that are in interim storage locations. What is the basis for this limitation? To reduce the potential economic loss if fuel damage occur I To minimize the potential for fuel damage if an assembly is droppe I To preclude the possibility of achieving criticality during a seismic even To preclude the possibility of an inadvertent process radiation monitor tri )

I QUESTION: 041 (1.00) i

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I During the insertion of a 't el assembly into a fuel storage rack the grapple ENGAGE / RELEASE switch is inadvertently p!a ed in the RELEASE position. What will occur? The grapple Jav/s will ope The grapple jaws will remain close ' The grapple release indicating light will illuminat The grapple jaws will open when the fuel assembly is grounde !

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LIMITED SENIOR REACTOR OPERATOR Page 26 of 30 i

QUESTION: 042 (1.00)

l A core reload is in progress on Unit 2. A fuel assembly has been grappled in the fuel pool and just raised to the NORMAL-UP position. The following occurs: I l -

Fuel pool level is recognized and confirmed to be LOWERIN The refuel floor ARM is NOT alarmin What IMMEDIATE operator actions are required? . Suspend movement of the fuel assembly at its present full-up conditio Place the fuel assembly in its assigned core location and stop moving fue Place the fuel assembly in the nearest open rack location and stop moving fue Notify the control room operator of decreasing water level. Wait for further directio i QUESTION: 043 (1.00)

Which of the below listed procedures require evacuation of the refuel floor in their immediate actions?

1) QCFHP 0110-02, inadvertent Criticality During Fuel Moves 2) QCFHP 0110-03, Fuel Assembly / Bundle Binding 3) QCFHP 0110-04, New/ Irradiated Fuel Damage 4) QCFHP 0110-05, Slow or Rapid Water Level Loss -

5) QCFHP 0110-07, Irradiated Fuel Above Normal Up Position ,2, and 3 ,2, and 5 , 3, and 4 ,4, and 5

LIMITED SENIOR REACTOR OPERATOR Page 27 of 30 QUESTION: 044 (1.00)

A fuel assembly was being lowered into the reactor core. As the bottom of the fuel assembly entered the top guide, hoist lower motion was automatically stopped and the following was observed:

- HOIST LOADED light OFF

- SLACK CABLE light ON

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fuel assembly is slightly tilted The grapple operator stopped fuel movement and informed the Fuel Handling Supervisor.

Permission to proceed was granted. In accordance with QCFHP 110-3, FUEL-ASSEMBLY / BUNDLE BINDING, what action should be taken? Slowly lower the hoist while shaking the mast to free the bindin Revise the NCTL. The fuel ast Embly needs to be returned to its original location in the fuel pool and released, Notify the reactor operator. Stop fuel movement. Lower an underwater camera to determine why there is bindin Raise Main Holst to load weight of fuel assembly onto the grapple. Reposition fuel assembly as needed to clear the Control Rod. Lower fuel assembly to its fully seated positio ,

QUESTION: 045 (1.00)

Using the guidance of the attached RWP 98-3004, if you are going to assist a fuel handler in removing a large known dose level item from the fuel pool, what requirements must be met?

_ Class 2 clothing with long gloves is acceptabl Class 4 clothing with a face shield are necessary; The ALARA briefing can be waived for this activity.,

' . An extremity dosimeter must be taped to your forear r ,1 1 l

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LIMITED SENIOR REACTOR OPERATOR Page 28 of 30 QUESTION: 046 (1.00)

A surveillance test of the Reactor Building Crane prior to operations in the Restricted Mode is required. Holst, Trolley and Bridge speed limits are to be tested and are required to be switched in slow speed. Which one of the following correctly describes the location of the speed selector switches? In the operating ca On the crane contro! pendan By the breakers located on the South Bridge walkwa In the control cabinets located on the bridge walkwa QUESTION: 047 (1.00)

When operating the Reactor Building overhead crane in RESTRICTED MODE, what is the speed of the bridge? feet / minut feet / minut feet / minut feet / minute.

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QUESTION: 048 (1.00)

Place the following major post-refueling activities in the best order of completion: Replace Reactor Head Install the Dryer Drain Rx cavity Install Fuel Pool Gates Install Drywell Head Install Shielding Blocks (cookies) Install Insulation Pack Install the Separator h, b, d, c, e, g, a, f h, b, d, c, a, g, e, f h, f, g, a, d, b, c, e h, a, f, e, d, g, b, c QUESTION: 049 (1.00)

Differential rod worth will peak at the: area of greatest thermal flux, because thermal neutrons cause fissio bottom of the core because all the fuelis exposed and able to cause fissio area of minimum fuel density because it has a greater effect on whether or not fission will occu top of the core because all of the fuclis covered and the greatest potential for fission exist <

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LIMITED SENIOR REACTOR OPERATOR Page 30 of 30 QUESTION: 050 (1.00)

What is the function of the REFUELING INTERLOCK CHECK pushbutton located in the

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901(2)-28 panel? It simulates a rod block for refueling interlock check It puts the refueling interlocks into active status when depresse It simulates the bridge over core condition for refueling interlock check This button simulates the one-rod-out condition for refueling interlocks check l

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LIMITED SENIOR REACTOR OPERATOR Page 1 of 6 EXAMINATION REFERENCES ANSWER: 001 (1.00) ANSVER: 007 (1.00) '

REFERENCE: REFERENCE:

UNIF801 QFP-600-2 Objectives S/UR/FH/B-0801-EK020 Objective UFH-0802-EK028 234000K504 ..(KA's) 234030K501 ..(KA's)

ANSWER: 002 (1.00) ANSWER: 008 (1.00) REFERENCE: REFERENCE:

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QCAP 0640-01 QCFHP 0400-06 step l NGET UF-0302-EK024 294001K103 ...(KA's) 201003K101 ..(KA's)

ANSWER: 003 (1.00) ANSWER: 009 (1.00) REFERENCE: REFERENCE:

S/UR/FH/B-0801-EK-022 UF-0805 233000K302 ..(KA's) Objective S/:/R/FH-0805-EK006 234000G2.2 ..(KA's) l AN5WER: 004 (1.00) ANSWER: 010 (1.00)

REFERENCE: OCAP 0630-06 REFERENCES:

NGET QCFHP 0400-06 294001K103 ..(KA's) Mods & LL 1997 Objective #1 ,

290002K305 ..(KA's) j ANSWER: 005 (1.00)  !

, ANSWER: 011 (1.00)  ;

REFERENCE: T/S 3.1 REFERENCE:

T/S 3. UF-0805, Objective S/UR/FH-0805-EK017 S/L 0805-EK021 234000K101 ..(KA's)

212000K407 ..(KA's)

ANSWER: 012 (1.00)

l ANSWER: 006 (1.00) REFERENCE:

REFERENCE: OCFHP 0100-05 UF-0805 S/UR-0805-EK020 UFH-0802-EK015 215004K402 ..(KA's) 234000K103 ..(KA's)

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LIMITED SENIOR REACTOR OPERATOR Page 2 of 6 ANSWER: 013 (1.00) ANSWER: 019 (1.00) REFERENCE: REFERENCE:

FH-0802 S/UFH-0803-EK005 Objectives UFH-0802-EK021h QCFHP 0100-01 234000K502 ..(KA's) 234000K401 ..(KA's)

ANSWER: 014 (1.00) ANSWER: 020 (1.00) REFERENCE: REFERENCE:

UF-0803 S/UFH-0803-EK013 Objectives S/UFH-0803-EK005 QCFHP 0100-01  !

UF-0280-lK009 234000K402 .(KA's) {

234000K502 ..(KA's) j l

ANSWER: 021 (1.00)

ANSWER: 015 (1.00) REFERENCE:

REFERENCE: S/UFH-0801-EK028 SYS-7500KO-03 00 QCFHP 0100-05 Objective S/UR/FH/FV-0805-EK011 234000K403 ..(KA's)

234000K403 ..(KA's)

l ANSWER: 022 (1.00)

ANSWER: 016 (1.00) d, REFERENCE:

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REFERENCE: S/UFH-0801-EK028 UF-0280-EK009 UFSAR 9.1.4. ,

234000K106 ..(KA's) 234000K404 ..(KA's) l ANSWER: 017 (1.00) ANSWER: 023 (1.00)

l REFERENCE: REFERENCE:

S/UR/FH-0805-EK016 UF-0805 ,

QCFHP 0110-02 Objectives S/UFH/FV-0805-EK008 !

234000K303 ..(KA's) 234000G2.2 ..(KA's) l

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ANSWER: 018 (1.00) ANSWER: 024 (1.00) REFERENCE: REFERENCE:

UFH-0804-EK015 S/UFH-0803-EK026 234000K105 ..(KA's) 234000K502 ..(KA's)

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LIMITED SENIOR REACTOR OPERATOR Page 3 of 6 ANSWER: 025 (1.00) ANSWER: 031 (1.00) < REFERENCE: REFERENCE: l S/UR/FH/B-0801-EK022 UF-0805 Objective S/UFH-0801-EK028 234000K503 ..(KA's) 234000K502 ..(KA's)

ANSWER: 032 (1.00)

ANSWER: 026 (1.00) ,

' REFERENCE:

REFERENCE: UF-0805 Objective: S/UFH-0805-EK016 UF-0803 Objective S/UFH-0803-EK005 234000G ..(KA's) l 234000K502 ..(KA's)

ANSWER: 033 (1.00)

ANSWER: 027 (1.00) i REFERENCE: 1 REFERENCE: S/UR/FH-0805-EK017 '

l LIC 0800 QCAP 1900-02 UFH-0800-EK005 QCFHP 0110-05 234000K505 ..(KA's) 234000A201 ..(KA's)

l ANSWER: 028 (1.00) ANSWER: 034 (1 00) REFERENCE: REFERENCE: ,

S/UFH-0803-EK015 UF-0805 234000K601 ..(KA's) Objective: S/UR/FH/FV/FC-0805.EK010 294001A110 ..(KA's)

ANSWER: 029 (1.00) ANSWER: 035 (1.00)

I REFERENCE: S/UFH-0803-EK020 REFERENCE:

234000K604 ..(KA's) S/L-0805-EK021 ,

T/S 3.10.A Action 2 I 234000G005 ..(KA's)

ANSWER: 030 (1.00) REFERENCE: ANSWER: 036 (1.00)

UF-0803 Objectives S/UFH-0803-EK013S UF-0280-EK009 REFERENCE:

234000K102 ..(KA's) S/L-0801-EK033 T/S Bases 3/4.1 G007 ..(KA's)

LIMITED SENIOR REACTOR OPERATOR Page 4 of 6 ANSWER: 037 (1.00) ANSWER: 043 (1.00) REFERENCE: REFERENCE:

S/L-0805-EK022 S/UR/FH-0805-EK017 T/S Bases 3/4.1 QCFHP 0110-02,03,04,05,07 234000G007 ..(KA's) 295023K301 ..(KA's)

ANSWER: 038 (1.00) ANSWER: 044 (1.00) REFERENCE: REFERENCE:

QCFHP 110-3 QCFHP 0110-03, Rev 0 Objective: S/UR/FH-0805-EK017 Objective: S/UR/FH-0805-EK017 295023G2.2 ..(KA's) 234000 ..(KA's)

ANSWER: 039 (1.00) ANSWER: 045 (1.00) REFERENCE: REFERENCE:

' UF-0805 NGET training objective Objective: S/UR-0805-EK019 RWP 98-3004 234000G2.2 ..(KA's) 295033K102 ..(KA's)

ANSWER: 040 (1.00) ANSWER: 046 (1.00) l REFERENCE: REFERENCE:

OCFHP 0200-02 FH-0802 Objective: UFH-0802-EK015 295023K102 ..(KA's) 234000K501 ..(KA's)

ANSWER: 041 (1.00) ANSWER: 047 (1.00) REFERENCE: REFERENCE:

S/UFH-0803-EK022 FH-0802 Objective: UFH-0802-EK014 295023K201 ..(KA's) 234000K501 ..(KA's)

ANSWER: 042 (1.00) ANSWER: 048 (1.00) REFERENCE: REFERENCE:

UF-0805 Objective: S/llR/FH-0805-EK003 QCFHP 0110-05, Rev. 2 234000K103 ..(KA's)

Objective: S/UR/FH-0805-EK016 233000A202 ..(KA's)

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LIMITED SENIOR REACTOR OPERATOR Page 5 of 6 ANSWER: 049 (1.00) ANSWER: 050 (1.00) REFERENCE: REFERENCE:

GP Reactor Theory Ch 5 Obj 8 UF-0803 Objective: S/UR-0803-EK016 Exam Bank #2687 234000A302 ..(KA's)

292005K104 ..(KA's)

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ANSWER KEY MULTIPLE CHOICE 001 c 021 c 041 b 002 b 022 b 042 c 003 b 023 a 043 d 004 c 024 b 044 d l

' 005 b 025 d 045 b 006 b 026 b 046 d 007 c 027 a 047 c 008 c 028 d 048 b 009 c 029 d 049 a 010 c 030 b 050 d 011 c 031 d ,

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012 d 032 c 013 b 033 d 014 a 034 a ,

015 d 035 d 016 d 036 c 017 . b 037 d 018 ~ b 038 c  ;

019 d 039 c 020 a 040 ' c ("""*"* END OF EXAMINATION *"*"**")