IR 05000354/1990017
| ML20197H731 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 11/02/1990 |
| From: | Conte R, Pullani S NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20197H711 | List: |
| References | |
| 50-354-90-17OL, NUDOCS 9011200099 | |
| Download: ML20197H731 (341) | |
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U.S. NUCLEAR REGULATORY COMMISSION REGION I ! OPERATOR LICENSING EXAMINATION REPORT l
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Examination Report No.: 50-354/90-17 (0L) i
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Facility Docket-No.: 50-354 l Facility License No.: NPF-57 I Licensee: Public Service Electric and Gas Company Post Office Box 236 Hancocks Bridge, New Jersey 08038 l Facility: Hope Creek Generating Station Examination Datesi August 29, 1990
. September 24-28, 1990 ,
October 1-3, 1990 !
Examiners: D. Draper, Examiner, PNL J. Muth, Examiner, PNL S. Pu11ani, Senior Operations Engineer C. Sisco,' Operations Engineer
. Chief Examiner: /du//an/ //- 02- 70 .!
'S. r Pullanl, y rerations Engineer Date l Approved by: .
// /Z /96 Richardfdnte, Chief 'Da ts i BWR SecTion, Operations Branch i Division of Reactor Safety Examination Summary: Written examinations and operating tests were administered !
to eight Reactor Operator (RO) candidates and nine Senior Reactor Operator 1 (SRO) candidates. Of the:nine SR0 candidates, only the written examination was ! administered to one SRO candidate who received a waiver for the. operating test
.and only the operating test was administered to another SR0 candidate who received a waiver for the written examination. All 17 candidates passed the '
examinations. Overall, the candidates were well prepared for the examination, d l Within this report, strengths and deficiencies were listed as feedback to the i licensed operator training program. During the course of the examination, a' i number of procedure deficiencies were noted for which the licensee initiated ; prompt corrective actions. 1 p 9011200099 901100 i PDR ADOCK 05000354 V PNU ,
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DETAILS !
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, rn t ~1. Introduction and 0verview
$M The NRC examiners administered.. initial examinations to eigh't R0' candidates
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p, and nine (including two who received waivers) SR0 candidates. The exami.- 1 nations were administered in'accordance with NUREG 1021, Examiner Stand ' l ]plI '
'ards, Revision-5. The .results of the examination are summarized below.
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_;- . . - 3l RO _ -l . SRO -l l- Pass / Fail l- Pass / Fail l
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:10perating I. 8/0- .1 :8/0 l i
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' .l0verall .l. 8/0 -- l 9/0 -l 1 . , ;l l l -l . ' - ,, ;
f, y2; : Persons Contacted' T l o , q
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n;2.1 Nuclear Regulatory' Commission (NRC)
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'S. . Dembek,= Acting Project Manager -(2),- (3)
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'd.: Canaday,' Examiner Trainee (2)-
sD.. Draper, Examiner,PNL(2),(3),-(4).. a JL Muth',. Examiner, PNL-(3), (4) 4-S.sPullani!,! Seni'o r Operations. Engineer (1)' (2),- (3), (4)
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J-C. Sisco, Operations: Engineer (1),-(2), (3),- (4)
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'2.2; Public Service Gas 'and Electric Company (FSE&G) ,
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' ~ , ~ Ci Bauer, Training Supervisor'(1),1(3)
f - D. Beckwith_,. Station._ Licensing Engineer (2) R
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R; Ceravolo,< Manager, Nuclear. Training-(4) _ _.
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LW. 'Gott,- Principal = Traini.ng Supervisor (1),' ~ (2), (4)- ,
~ R 'Hovey, Operations Manager. (2), (4)" .
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C. Johnson, General Manager, Hope Creek (Acting) .(2),. (4) ' j
. .S.- Jones, Training Supervisor-(1) . ' '
D. -Kabachinski, Simulator Operator <.(3), j M. Martin,L Training / Instructor (1)
.G. Mecchi,; Operatio_ns Training (2), (4)- 3
: G. :Munzenmaier,; General Manager,' Nuclear Serv _ ice's (4)'
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W. -O'Malley, -Operating Engineer (1), :(4) a D. Rein, Simulater Operator (3) A J. . Sparks, Senior' Nuclear Training Supervisor (3), (4) r
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Notes (1) Denotes those.present during the pre-examination review of the written examination on September 19, 1990 (2), Denotes those present during the entrance meeting on September 24, 1990 (3) Denotes those present duri g the Simulator Examination Validation and Verification on Septemver 25, 1990
' (4) Denotes those present'during the exit meeting on October 4, 1990 .--
3. Pre-Examination Activities
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3.1 License Application Review The license applications were reviewed.in accordance with NUREG 1021, Examiner Standards, Revision 5. The applications contained suf fi-
-cient-information to determine the eligibility of the applicants to appear for the examination or to grant-the waivers as requested.
3.2 Examination Preparation The written examination and operating test were prepared in accord-- ance with NUREG 1021, Examiner Standards, Revision 5. The reference material provided by the licensee was found adequate for the preparation.
3.3 Pre-Examination Review Prior to administration of the written examination, on September 19, 1990, the facility reviewed the examination at the Regional Office.
All facility comments were discussed and resolved during the review session. The examination was revised to incorporate those changes.
All-facility individuals involved with the review of the examination y materials signed security agreements to ensure that there was no-compromise of the examination.
3.4 Entrance Meeting An entrance meeting with.the licensee was held on September 24, 1990, at .its-training facility. The purpose of. the meating was to discuss the plan and schedule for the examination.
3.5 Simulator Examination Validation and Verification All simulator examination scenarios used for the operating test were validated and verified at the facility simulator on September 24, 1990. All facility individuals involved with the validation and
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verification signed security agreements to ensure that there was no compromise of the examination.
4. Examin'ation - Related Findings and Conclusions The following is a summary of the strengths and deficiencies noted during the administration of the written exam. nation and operating tests. This information is being provided to aid the licensee in upgrading license -, and requalification training programs. No licensee written response is ! required. l
The overall assessment by the examination team was that the applicants. .
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i were well prepared for the examination. l 4.1 Written Examination There were no specific strengths or deficiencies noted during the - administration of the written examination.
4.2 Operating Test Strengths
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The applicants,.in general, demonstrated good knowledge in the j administrative topics. i l
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Both SRO-and RO. applicants-demonstrated good knowledge'and a familiarity with plant systems and components, their instru ' a
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mentation and controls, associated procedures'and administrative requirements except the specific deficiency identified below. .q
] * . The applicants, in general., demonstrated good 'understar: ding and' ;
interpretation of annunciators'and alarm signals. '
: *. The applicants, in general, demonstrated their ability to H quickly and accurately diagnose the events or conditions based on signals or.other instrument readings. This observation was j '
especially-evident in the case of the SRO applicants. l,
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The applicants compliance and usa of procedures were found- [ satisfactory except as identifiea below under weaknesses. The l SRO. applicants were quick to ent" the E0Ps based on entry ,* conditions.
The applicants, in general, demonstrated satisfactory perform-ance in control. board operations except the specific deficiency .{ ; identified below.
The communication and crew interactions during the simulator exam was generally satisfactory. The SRO applicants briefed i their crews periodically during the scenarios.
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's The SR0 applicants, in general, were able'to direct shift ,
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La - The applicants, in general, demonstrated good knowledge and } familiarity in the compliance and use of Technical .) Specifications. ,7
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Deficiencies q
~* Several applicants were not able to demonstrate their fami. liar- 3
' 0'ity to reset a recirculation pump runback. ,J U -4 Some; applicants did not reopen the recirculation pump discharge
-. t valves.af ter 5 minutes, as required by the system-operating 1 1 procedure. They.all did close-the discharge valves after the- *
pump . trip', as required by the procedure. 3 4i3'LAdditional Findings , ,
.l iDuring?the1 preparation >and administration of operating tests,7the.NRC I i nation team:identifiedcthe following plarit procedure
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, ;11 - Administrative Procedure HC.0P-OL-72.0026(Q), HCGS Surveilla'n'ce - -
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- Log - (Control.' Room)',' Rev'i sion 30, MStep . 3 L11.1'.31, incorrectly '
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; stated that MET Data is.available:in the! control room from the' !
RMS. computer. :
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2. i Alarm Respo'nse Procedure HC.0P-AR-ZZ.0011(Q), Mn 'Stm Line
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Radiation HI, Revision 11,Ldid'not provide;d$rection for
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, obtainingLa; key'necessary:to place'an. inoperable Mai.n: Steam
!:- g e ;Line: Rad i Monitor .in the tripped condition.
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3. -Alarm Response Procedure HC.0P-AR-ZZ.014(Q),l Turbine High. Vibra- !
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s -tripping-the turbine at-:12-Mils.
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' _4 . :-System Operating Procedure _HC~.0P-S0-AC.001, Main TurbinecOpera--
tion, did not provide' direction for operatorLactions prior to ' tripping the-turbine at 12 Mils.
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, 5. - : Abnormal Operating. Procedure HC 0P-AB.ZZ-121(Q), Failed Open 1
~ Safety / Relief Valve, did-not provideLprocedural. transition-into:
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. the. Emergency: Operating Procedures (EOPs)_ at 95 F in the: a; suppression-pool.
- = 6. . Integrated _-Operating Procedure HC.0P-IO.ZZ-008(Q), Shutdown Outside:-the Control Room,' Step 5.1.11-(Caution), ir.cortectly stated <800 psig as the minimum pressure to open a Safety Relief Valve. The correct value is <700 psig to agree with the-
.EOPs.
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6 1 7. System Operating Procedure HC.0P-SO.BD-001(Q), Reactor Core Isolation Cooling (RCIC) System Operation, Step 5.5.9, did not prevent the RCIC to FW isolation and the RCIC full flow test return to the CST valves to be open simultaneously during trans-fer from the full flow test mode to RPV injection. A potential existed for hot water to reach the CST and flash to steam i creating a water hammer within the CST.
< 8. Inservice Test Procedure HC.0P-IS.BD-001(Q), Reactor Core
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Isolation Cooling Pump Inservice Test, Step 5.1.7, incorrectly stated that HV-F022 was to be throttled open 15 seconds. The correct value is.7-8 seconds in accordance with System Operating Procedure HC.0P-SO.BD-001, Step 5.5.3.
9. System Operating Procedure HC.0P-SO.KE-001(Q), Refueling . 1 Platform and Fuel Grapple Operation, did not clearly state that - a SR0 licensed or SR0 limited to fuel ' handling operator is . required to supervise all irradiated fuel movement on the .I refuel floor,
, j Tne licensee took prompt corrective action by initiating appropriate'
procedure change request for each of the identified procedure defi- l ciencies. The examination team established confidence in~the licen- J see's planned corrective action te assure proper resolution of the ! noted deficiencies. The area of procedure adequacy will continue to l be routinely reviewed by NRC staff. j 5. Exit Meeting An exit meeting was conducted on October 4, 1990, following the admini-stration of the examinations. The licensee representatives that attended 4
.the meeting are listed in'Section 2. of this. report. The. licensee presented a -few additional comments on the written examination , questions j that were not identified during the pre-examination review (see. Attachment.
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'The generic strengths and deficiencies noted on the operating examinations *
were presented (see Section 4.2 of this report). The additional findings i by the examiners while administering the operating tests (see Section-4.3) , were discussed. The chief examiner stated that the results of the exami- ' nations would not be presented at the exit meeting, but would be contained I in the examination' report and that every effort would be made to send-the l results in approximately 30 working days, l
. Attachments- l 1. Reactor Operator Examination and Answer Key
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=i 2. Senior Reactor Operator Examination and Answer Key 3. Facility Comments on Written Examination 4. 'NRC Resolution and Facility Comments '
5. Simulator Facility Report i a
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' Reactor Operator Examination and Answer Key- ,
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NRC 0fficial Use Only
i Nuclear Regulatory Commission ariint$o
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i This document is removed from Official Use-Only category on
.date of examination.
NRC Official Use Only l, L
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'U.'S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION REGION 1 FACILITY: Hope Creek 1
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REACTOR TYPE: BWR-GE4
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DATE ADMINISTERED: 90/09/24 CANDIDATE:
INSTRUCTIONS TO CANDIDATE: .
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Points for each question are indicated in parentheses after the question. .To pass this examination, you must achieve an overall grade of at least 80%. Examination papers will be picked up four and one half (41/2) hours after the examination starts.
- NUMBER TOTAL CANDIDATE'S CANDIDATE'S i 1 QUESTIONS POINTS' POINTS OVERALL _ t GRADE (%)- ;
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U 100 '100.00 l L,
i o All work done on this-examination is my own. I have neither given nor received aid. - L Candidate's Signature-
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NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIJNS t During.the administration of this examination the following rules apply: 1
:1 Cheating on the examination means an automatic denial of your application and could result in more severe penalties. -
2. After the exanonation'has been completed, you must sign the statement on
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the cover sheet indicating that the work is your own and you have not . received or_given assistance in completing the examination. This must be i done after you complete the examination. '
.3. 'Restroom trips areLto be limited. and only one candidate at a time may leave. You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.
.4. Use black ink or dark pencil only to facilitate legible reproductions.
5. Print your name in the blank provided in the upper right-hand corner of the exam;.ation cover sheet.
6. Fill in the date on the cover sheet of the examination (if necessary).
L7. You may write your answers on the examination question page or on a separate sheet of paper. USE ONLY THE PAPER PROVIDED AND DO NOT WRITE ON
' 'iHE BACK SIDE.0F.THE PAGE.-
8.- If you write your answers on' the examination question page and you need more; space to answer a specific question, use a separate sheet of the paper provided and insert .it directly after the specific question. DO NOT WRITE ON THE BACK SIDE OF THE EXAMINATION QUESTION PAGE.
9; Print your name in the u)per right-hand corner of the first page of answer
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' sheets whether you use t;e examination question pages or separate sheets of paper. Initial each'of the following answer pages.
10.;Before you turn in your examination, consecutively number each answer ' sheet, including' any additional pages-inserted when writing your answers
:on the examination _ question page..-
11. If you are using separate sheets, number each answer and skip at least 3 lines between answers to allow space for grading.
-12. Write "Last Page" on the.last' answer sheet,
- i 13. Use abbreviations only if they are commonly used in facility literature.
-Avoid usi_ng symbols such as < or > signs to avoid a simple transposition
: error resulting.in an incorrect answer. Write it out, t -
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.__ [14.EThepointvalueforeachquestionisindicatedinparenthesesafterthe question. :The amount of blank space on an examination question page is NOT an indication of the depth of answer required. '
15~. Show all calculations, methods, or assumptions used to obtain an answer.
- 16'E Partial credit may be given. Therefore, ANSWER ALL PARTS OF'THE QUESTION
:AND DO.N0T LEAVE ANY ANSWER BLANK. NOTE: partial credit will NOT be given.on multiple ch'oice questions.
..17. Propo'rtional grading will be applied.. Any additional wrong information
. that is provided may count against you. For example, if a question is worth one point and asks for four responses, each of which is worth 0.25 points, and you' give five responses, each of your responses will te worth 0.20 points. If one oJ vaur five responses is incorrect, 0.20 will be.
. deducted and your total redit for that question will be 0.80 instead of
'1.00 even though you got the four correct answers.
. 18.' If the intent of a question is unclear, ask questions of the-examiner only. ' 19. When turning in your examination, assemble the completed examination with examination questions, examination aids and answer sheets. In addition, turn in all scrap paper.
20. To pass the. examination, you must achieve an overall grade of 80% or greater. , 21. There is :a time limit of (41/2) hours for completion of the examination.
. (or 'some other time if- less than the~ full examination-is taken.)
P 22. When you are-done and'have turned in your examination,. leave the examin- , Lation area'as defined by the examiner. If you are found in this area
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REACTOR OPERATOR Page. 4
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QUESTION: 001 (1. 00)' Vsing the Attachment-#1 drawing of the primary containment,
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o. ,a. Biological Shield b. Pedestal
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c. Drywell ;
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d. Pressure' suppression chamber
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Consider the following conditions: Turning-gear selected for AUTO mode Main turbine rotor " stopped" (as sensed by low speed stop switch) One main turbine lube oil lift pump running WHICH ONE (1) set of conditions will automatically start the turning gear piggy back motor? a. Turbine speed less~than 100 rpm (as sensed by magnetic pickups in the quill shaft failure circuit), main generator disconnect open, normal main turbine oil. supply pressure.
- b. Normal' main turbine bearing oil pressure, both main generator output breakers open, main generator disconnect open.
c. :Both main generator _-output breakers open, main generator
. disconnect open, turbine speed less than 100 rpm (as sensed by-magnetic speed pickups in the quill shaft failure circuit)
d. Normal . main turbine bearing oil' supply pressure, both main
- generator output breakers' open, two main turbine. lube oil lift pumps-running.
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QUESTION: 003- (1.00)
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lWHICH ONE (1) of the following, conditions will- yield an immediate
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o .i a. Lube oil pressure of 25 psig -
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b. High RPV pressure of 1071 psig-
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c. Hoisture . removal valve (HV-1765) not 100% open d. Main condenser pressure of 6.6" Hg abs. ,
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+ REACTOR OPERATOR. Page 7 , -QUESTION: 004 -(1. 00) ' :The Resitual Heat Remcval (RHR) System is operating in the Shutdown ; Cooling ~(SDC) Mode. WHICH ONE (1) of the followin) system responses 'will occur if RPV pressure were to increase to 90 psig? '
s a '. Shutdown cooling suction valve (F008) auto closes, RHR pump suction valve (F006 A/B) auto closes and outboard head spray valve (F023) remains open-
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s b. - All running RHR pumps trip, RHR pump suction valve (F006 A/B) '
-auto closes, outboard head spray valve (F023) remains open i
Lc. . Shutdown cooling suction valve (F008) auto closes, all running l RHR pumps trip, outboard head spray valve (F023) closes d. Shutdown cooling suction valve .(F008) auto clnses, all running ) RHR pumps trip, RHR pump suction valve (F006 A/B) auto closes ,-
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I REACTOR OPERATOR Page 8 i g QUESTION: 005 (1.00) 1) of the following conditions will cause a FULL runback of I WHICH the recircuONE (lation pumps? i a. Secondary condensate pump trip with FW flow > 85% I b. Primary condensate pump trip with FW flow > 85% ! l c. An RFPT trip (3 running) with RPV level < 30" d. Loss of one circ water pump (3 running) and condenser pressure is 4.5"hg
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QUESTION: 006 (1.00) I to110 wing a required initiation of the Standby Liquid Control system, you are directed by the " Level / power Control" procedure (0P-E0-ZZ-207) to * Lower RPV water level by terminating and preventing all injection except from CRD and boron injection systems until ....." WHICH ONE < _1)
( of the following conditions are those that allow coolant injection to be re-initiated? ,
t a. RPV water level reaches TAF, all SRVs remain closed, drywell pressure remains below 1.68 psig and SLC tank volume reaches 5% b. Reactor power drops below 4%, all SRVs remain closed, drywell
. pressure remains below 1.68 psig, and SLC tank volume reaches 5%
c. Reactor power drops below 4%, RPV water level reaches TAF, all SRVs remain closed, and SLC tank volume reaches 5% d. Reactor power drops below 4%, RPV water lev 21 reaches TAF, all SRVs remain closed, drywell pressure remains below 1.68 psig
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I REACTOR OPERATOR Page 10 I QUESTION: 007 (1.00) A fire occurs in the Control Room requiring immediate control room evacuation. HPCI and RCIC auto initiated prior to establishing control l at the shutdown panel. At the shutdown panel, all channel transfer switches are turned to "EMER." WHICH ONE (1) of the following describes the response of HPCI and RCIC systems?
-a. Both HPCI and RCIC continue operating b. HPCI continues operating while RCIC trips ,c. IlPCI trips and RCIC continues operating d. HPCI and RCIC both trip i
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REACTOR OPERATOR Page 11 ! \ QUESTION: 008 (1.00) WHICH ONE (1) of following is the purpose of the positive vent valves of the scram pilot valves in the control rod drive hydraulic system? a. To ra3 idly interrupt the air supply and provide a vent path for tie rapid closure of the scram discharge volume vent and drain valves, b. To direct air pressure to the scram inlet and outlet valves, holding the scram valves closed.
c. To provide a redundant means of bleeding air from the scram inlet / outlet valves, d. To minimize the air exhaust time from the scram valve air header resulting in quick scram valve opening.
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i REACTOR OPERATOR Page 12 QUESTION: 009 (1.00) WHICH ONE (1) of the following. signals listed below will cause a direct recirculation drive motor breaker trip? l l a. Loss of stator cooling (3-5 minute time delay) I
b. High drywell pressure of 1.68 psig i c. Both M/G ventilatio:. fans trip i d. Loss of one (1) cire water pump (3 running) and main condenser pressure of 4.3" Hg
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QUESTION: 010 (1.00) WHICH ONE (1) of the following items will bypass the APRM Upscale Thermal Power trip for the reactor protection system (RPS)? a. Mode switch in Run , b. APRM bypass switch in Bypass c. Mode switch in Startup d. IRM range switch on Range 1
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REACTOR OPERATOR Page 14
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' QUESTION: 011 (1.00) A tour group is currently visiting Hope Creek during your shift. One of the group is a 36 year old. This individual is an ex-radiation worker, and has completed an NRC-4 form with a total exposure of 88 REM.' This individual has completed the GET and RWT courses and as such his exposure is not limited due to his " visitor" status.
For this individual, WHICH ONE (1) of the following would be the administrative quarterly dose limits? (no extensions) a. 10 mrem /qtr b. 300 mrem /qtr c. 1000 mrem /qtr
.d. 2000 mrem /qtr i
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QUESTION: 012 (1.00) In calculating the APRM flow biased thermal power upscale scram trip setpoint, a factor "T" must be determined. WHICH ONE (1) of the following states the definition for "T7" a. Average Planar Linear Heat Generation Rate / Linear Heat Generation Rate b. Fraction of Rated Thermal Power / Core Maximum r eaction of Limiting Power Density c. Actual Core Power / Critical Power for Acts.a1 Core flow d. Linear Heat Generation Rate /Fractica of Rated Thermal Power
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. REACTOR OPERATOR Page 16 QUESTION: 013- (1.00)
- Given the following control rod directional control valve sequence: (DCV) 123 and (DCV) 121 energize for approximately 2.0 seconds and then de-energize, then (DCV) 120 energizes for approximately-le5 seconds. WHICH ONE (1) of the following describes the Rod motion that results?
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a. Continuous insertion b. Continuous withdrawal c. Single notch insertion , d. Single notch withdrawal
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REACTOR OPERATOR Page 17 l QUESTION: 014 (1.00) WHICH ONE (1) of the following describes the scram discharge volume (SDV) vent and drain valves responce during a reactor scram? a. Solenoid operated valves energize to remove control air from the operators of the vent and drain valves, closing them.
b. Solenoid operated valves de-energize to remove control air from tie operators of the vent and drain valves, closing them, c. Solenoid operated valves energize to supply control air to the operators of the vent and drain valves, closing them, d. Solenoid operated valves de-energize to supply control air to the operators of the vent and drain valves, closing them.
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REACTOR OPERA 10R Page 18 QUESTION: 015 (1.00) WHICH ONE (1) of the following actions will open the main turbine bypass valves after a turbine trip?
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a. decrease load limit pot b, decrease maximum combined flow pot i c. decrease pressure setpoint d. decrease steam throttle pressure l l
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f REACTOR OPERATOR Page 19 QUESTION: 016 (1.00) The output of both EHC pressure regulators fail high causing a bypass valve to open. WHICH ONE (1) of the following EHC controls can be used to close the bypass valve? a. Bypass jack b. Load limit c. Pressure set d. Maximum combined flow limiter
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- REACTOR OPERATOR Page 20 QUESTION: 017 (1.00)
Given the following heat up data: TIME RPV Pressure 7:00 am 233 psig 7:15 am 233 psig 7:30 am 233 psig 7:45 am 233 psig 8:00 am 233 psig 8:15 am 408 psig WHICH ONE (1) of the following -is the latest time at which heatup must be secured in order to prevent exceeding the Technical Specification limit for heatup at the current heat up rate? a. 08:30 am t b. 08:45 am c. 09:00 am d. 09:15 am
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REACTOR OPERATOR Page 21 I i I i QUESTION: 018 (1.00) .
During a core traverse by the TIP system with the probe in the i core, conditions indicating a loss of Coolant Accident (LOCA) l occur causing a Group XVII isolation. WHICH ONE (1) of the i following statements describes the response of the TIP system? l
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a. It continues its sequence for the present core position I and then reverses and withdraws into the shield. I b. The ball valve closes regardless of the position of the probe, shearing the probe and isolating the TIP system.
c. The TIP system reverses, slow if in the core and fast if outside the core, withdraws into the shield and the ball valve closes.
d. The TIP system reverses and moves out in fast speed, regardless of position, into.the shield and the ball valve closes.
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_ _ . REACTOR OPERATOR Page 22 ) l l QUESTION: 019 (1.00)
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A tain generator hydrogen leak has occurred. Hydrogen pressure I has stabilized at 45 psig. The plant is operating at a 0.95 ) leading power factor. WHICH ONE (1) of the following is the main generator power output limit based on these conditions? (Use Attachment #2) a. 1000 MW b. 1050 MW c. 1100 MW d. 1150 MW
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, L REACTOR OPERATOR Page 23 QUESTION: 020 (1.00) During power operations with the reactor at 782 MWe, the load dispatcher requests that you decrease power to 223 MWe as quickly as possible. WHICH ONE (1) of the following is the minimum time in cinutes required for this downpower? (Use Attachment #3) a. 2 min b.' 9 min c. 10 min d. 20 min L c ! - J
REACTOR OPERATOR Page 24
QUESTION: 02) (1.00) Given the following power history:
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0800 1000 101 %
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1000 1100 98 %
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1100-1300 100.5%
-- 1300 1500 99 %
WHICH ONE (1) of the following is the maximum percent of power allowed over the next hour without violating Tech Specs? a. 99% b. 100% co- 101% d. 102%
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QUESTION: 022 (1.00) With the reactor at 100% power, WHICH ONE (1) of the > following conditions could cause a control rod to drift OUl? a. DCV 120 fails open (settle / direction control valve) b. HCU accumulator depressurizes c. Collet fingers stuck out d. Buffer piston clogged
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' REACTOR OPERATOR Page 26 QUESTION: 023 (1.00) WHICH ONE (1) of the following selections describes the flow path order for the offgas system? a. Hold up pipe, strip heaters, gas recombiner preheater gas recombiners, gas cooler condenser, glycol cooler condenser, guard bed, adsorber bed, HEPA filters, stack b. Gas recombiner prehea' strip heaters, gas recombiners, gas cooler condenser, hole rp pipe, glycol cooler condenser, guard bed, adsorbcr bea, HEPA filters, stack c. Gas cooler condenser, glycol cooler condenser, gas recombiner preheater, strip heaters, gas recombiners, hold up pipe, guard bed, adsorber bed, HEPA filters, stack d. Guard bed, strip heaters, gas recombiner preheater, gas recombiners, gas cooler condenser, hold up pipe, glycol cooler condenser, adsorber bed, HEPA filter, stack
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-QUESTION: 024- (1.00) !
WHICH ONE (1) of the following is an assumption for negative reactivity . addition used in the design of the standby liquid control system? !
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a. 10% shutdown margin b. Iodine decay c. Xenon decay .i i-c d. Samarium decay
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REACTOR OPERATOR Page 28 1
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QUESTION: 025 (1.00) WHICH ONE (1) of the following is the pressure boundary for the IRM detector assembly? a. Shuttle tube b. Dry tube
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c. Guide tube d. Drive tube
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REACTOR 0PERATOR Page 29 r F QUESTION: 026 (1.00) , WHICH ONE (1) of the following components of a control rod provide a path for. reactor water to initiate a scram to a full out (position 48) rod when no accumulator pressure is available? a. Stop piston b. Valve disk c. Piston tube orifices d. Collet' assembly ! ' r
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REACTOR OPERATOR Page 30 ! i QUESTION: 027 (1.00) . WHICH ONE (1) of the following describes the location of the pressure . tap in the CRDH system for the control room indication of charging header I water pressure? l a. Between the flow control station and the drive water l pressure control valve l b. Between the drive water filters and the flow control station c. Between.the pressure control valve and the pressure equalizing valves d. Between the CRD pumps and the drive water filters
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REACTOR OPERATOR Page 31 QUESTION: 028 (1.00) WHICH ONE (1) of the following describes the reset function of the RCIC turbine? a. The RCIC mechanical overspeed trip can be reset either locally or in the control room.
b. The RCIC initiation logic " Reset" pushbutton resets ! both the manual and automatic initiation signals.
c. The RCIC mechanical overspeed trip can be reset from the , control room only. I d. Resetting the trip throttle valve is required for all turbine trips that close the Trip Throttle valve HV-4282 prior to restarting RCIC turbine.
- - . . -. _ _ . REACTOR OPERATOR Page 32 QUESTION: 029 (1.00) WHICH ONE (1) of the following control room annunciators will occur as a direct result of a main generator excessive volt /hz condition? a. Power system stabilizer trip
.b. Unit protection lockout relay trip c. Exciter trip to DC regulator d. Generator field ground i
REACTOR OPERATOR Page 33 : QUESTION: 030 (1.00) WHICH ONE (1) of the following conditions will cause a trip on the main turbine? a. ETS fluid pressure of 1100 psig at the front standard b. Speed 105% of rated c. Exhaust hood temperature of 180F d. Low shaft oil pump discharge pressure with turbine > 1300 RPM
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REACTOR OPERATOR Page 34 l l QUESTION:-031 (1.00) WHICH ONE (1) of the following conditions allows rod motion? a. 7 LPRM inputs to the RBM b. Switch in calibrate c. More than one rod selected d. Failure to null l t i l i j' i
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QUESTION: 032 (1.00) WHICH ONE (1) of the following is designed to mitigate the consequences of a control rod eject accident? a. Rod Worth Minimizer + b. Rod Block Monitor c. Reactor Manual Control d. Control rod drive support housing
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REACTOR'0PERATOR Page 36 l, J QUESTION:.033 (1.00)- ,
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On a' design basis LOCA, certain reactor components ensure two thirde ' coverage is maintained. WHICH ONE (1) of the following are NOT one of these components? a. Shroud support ring b'.- l Jet pump diffuser and mixer c. Core shroud ; d, Core support plate , i
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. QUESTION:.034 (1.00) .
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.During a startup following a refueling outage, WHICH ONE (1) of the '
t following~ is the maximum allowable difference between actual rod density and the predicted rod density? ! a. 0.1% Delta K/K i b. 0.5% Delta K/K c.1.0% Delta K/K d. 10.0% Delta K/K : ,
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QUESTION: 035 (1.00) WHICH ONE (1) of the following intermediate range monitor rod blocks is ' bypassed when the range selector is positioned to RANGE I? a. IRM Hi 3 I b. IRM Downscale c. IRM INOP i d. Detector Not Full In ,
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'QUESTIONi 036- (1.00)
WHICH ONE (1) of the following conditions will shutdown a diesel generator following an emergency start? a. Reverse power b. Underfrequency c. Bus differential overcurrent d; Overvoltage
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REACTOR OPERATOR OVESTION: .037 (l.00) WHICH ONE (1) of the following describes the control rod drive hydraulic system response during a scram? a. The scram pilot valve energizes to vent the air off of .the scram inlet and outlet valves.
b. The scram pilot valve energizes to vent the air off of the scram discharge volume (SDV) vent and drain valves.
c. The back up scram valves de-energizes to vent the air off o. the scram inlet and outlet valves, d. The backup scram valves energize to vent air off the scram inlet and outlet valves.
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t QUESTION: 038 (1.00)' i
.WHICH ONE (1) of the following selections will cause a Group ic MSIV sealing: system isolation? +
t a. Reactor bldg exhaust 1.3 EE -4 uCi/cc i b. Drywell pressure 1.60 psig c. Refuel floor exhaust 1.5 EE -3 uCi d. Reactor vessel level -38 inches t l- . l l
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QUESTION: 039 (1.00) WHICH ONE (1) of the following statements describes the shorting i links that are used in reactor protection system (RPS)?
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Installation of the shorting links activates the SRM
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Scrams and bypasses the IRM and APRM scrams.
' b. Removal of the shorting links activates the SRM scrams in a coincidence of one-out-of-two-twice logic scheme. . a c. Installation of the shorting links activates the SRM, < IRM and APRM scrams in one-out-of-two-twice logic schemes, d. ' Removal of the shorting links enables a scram if any single SRM, IRM or APRM channel trip.
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. REACTOR OPERATOR- Page 43 QUESTION: 040' (1.00)
i WHICH ONE (1)-of.the following SRM rod blocks / scrams art bypassed with-its I associated IRMs on Range 3? e
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p 'a. SRM ' Inoperable
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b.-SRM-detector wrong position i c. SRM upscale reading 1.5 EE +5
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d. SRM detector not full in . 2 [ ,, %- D V pg , a
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' QUESTION: 041 (1.00)
l During a startup with the reactor at 10% power, the rod sequence l control ~ system becomes inoperable. In accordance with Technical Specifications, WHICH ONE (1) of the-following actions must be taken? a. All rod movement must be- second verified in order for rod withdrawal to continue.
b. Control rod movement shall not be permitted except by scram.
c.. Commence a unit shutdown by inserting control rods in ; accordance with the approved sequence package,
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d Rod withdrawal may continue provided the RWM is operable.
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REACTOR OPERATOR Page 45' QUESTION: 042 (1.00) k'HICH ONE (1) of the following would NOT occur as an automatic response to a valid high drywell pressure signal? . a. Instrument gas TIP purge isolation < b. : Recirculation loop B sample line isolation c. Torus water cleanup / suppression. pool' supply inboard isolation d. . Containment hydrogen / oxygen analyzer suction isolation -
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REACTOR OPERATOR Page 46 l l l QUESTION: 043- (1.00)
> 0P-EO.ZZ-201 " Alternate Level Control" has been eatered. RPV levei has l -
l dropped'to -200 inches on'the Fuel Zone indication. The procedure requires
. emergency depressurization. WHICH ONE (1) of the following is the reason for '
depressurization? ' a. Maintain peak cladding temperature below 1800F b. Maintain total oxidation of the cladding less than 0.17 of the t total cladding thickness c. Maintain the maximum hydrogen generation less than.0.01 times the hypothetical maximum ; i d. Maintain a coolable geometry for long term cooling '
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QUESTION: 044 (1.00)- GIVEN that: /
- Load selector - 1067 MWe - Load limit = 107 % - Max. comb. limit - 113% - Throttle pres. - 947 psi _g - 'A' pressure reg =_920 psig ' ,i The 'A' EHC pressure regulator setpoint fpfis downscale. (see Attachment #4.
point #1) WHICH 0NE (1) of the followin 'is the resulting bypass valve steam
- flow signal of the EHC logic system? -i l .
a. 0% steam flow l
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l- ' b. 6% steam flow ; c. - 7% steam flow [ 1 i d. 13% steam flou 7 4
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QUESTION: 045 (1.00) WHICH ONE (1) of the following states the function of the Intercept valves (IV's)? a. Provide backup protection for the main turbine generator if the main turbine stop valves fail b. To direct steam to the fifth point feedwater heaters c. To provide temperature control during shell warming d. To prevent turbine overspeed upon loss of generator load i-l t
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QUESTION: 046 (1.00) Intaccordance with FSAR and " Containment Hydrogen Recombiner System" WHICH ONE (1) of the following is the MAXIMUM time allowed before system initiation is required after a LOCA event? a. Within 8 hours b. Within 12 hours < c. Within 24 hours d. Within 34 hours , i 1-
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'QUPSTION: 047 (1.00) 'Due to vacations and required training, you have been asked ,
to work the following schedule next week.
Sunday 0800-2000 Monday 0800-2000 Tuesday 0800-2100
' Wednesday 0800-1700 Thursday 0800-1600 '.
Friday 0800-2100 Saturday NONE WHICH ONE (1) of overtime guidelines, as outlined in SA-AP.ZZ-002
" Station Organization and Operating Practices," would be violated.
Assume times DO NOT include TURNOVER and the plant-is ope ating at rated power.
a. No more than 16 hours in a 24 hour period b. No more than 24 hours in a 48 hour period , c. At least 8 hours rest between work periods- ,
:d. No more than'72 hours in any 7 day period ,
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QUESTION: 048 (1.00) , l A; Technical Specification Quarterly (92 days) surveillance , was performed on the High Pressure Coolant Injection system on JUNE 10, 1989, SEPT 25, 1989 and JAN'1, 1990. WHICH ONE (1) of the following is the last MONTH and DAY the next quarterly , surveillance can be performed on? (Use Attachment #7)
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q a. April 1,.1990 b. April 5, 1990 c. April 25, 1990 d. April 30, 1990 i
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. QUESTION: 049' (1.00)
WHICH ONE (1) of the.following is an UNNECESSARY condition for-maintaining secondary containment integrity? a. All secondary containment hatches and blowout panels are closed and sealed b. For double door arrangements, at least one door in each access to_the secondary containment is closed
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c. The sealing mechanism associated with each secondary i containment penetration is operable d. Maintain the suppression chamber between an indicated L
' level of.74.5" and 78.5". <
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QUESTION:-- 050 (1.00) ,y During.-an ATWS accident, you are injecting boron using the SBLC pumps. WHICH'ONE (1) of the following tank levels equates to when
' COLD SHUTOOWN BORON WEIGHT" is reached?
a.1800 gals
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b.1600 gals , c. 1400 gals d.1200 gals
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g QUESTION: 051~ (1.00): ;
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- WHICH'ONE-(1)-- ofLthe. following group isolations occur as a- q
: result"of a main steam line radiation high-high?
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a. Group 2 Rx recirc water sample system j
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c.; Group 9 Drywell sumps-
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d. Group 12 Containment atmosphere control. '; i
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QUESTION: 052 (1.00) WHICH ONE (1) of the following group isolation will NOT occur as a result.of a drywell pressure high of 1.68 psig? l a. Group-Ic MSIV sealing system isolation valves i b. Group 2 Rx recirc water sample system c. Group 13 Hydrogen /0xygen analyzer system d. Group 17 Transversing in core probe
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QUESTION: 053 (1.00) WHICH ONE (1) of the following conditions require an ! entry.into E0P-102A, " Primary Containment Control?" a.. Drywell temp 134 F i b. Suppression pool level 74 inches ; c. Drywell Pressure 1.2 psig.
[ d. Primary containment Hydrogen 1.68% <
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- QUESTION- 054 _(1.00)
The~ function of the end of cycle RPT is to prevent exceeding WHICH ONE
'(1):of the following limits?
a. MCPR (minimum critical power ratio) !
-b. MAPLHGR (maximum average planar linear heat generation i rate)
c. CMFLPD (core maximum fraction of-the limiting power ! density).
d. FRTP (fraction of rated thermal power) J i l
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-REACTOR OPERATOR Page 58 QUEST 10N: 055 (1.00)
WHICH ONE.(1) one of the following will prevent initiation of the MSIV: sealing: system? .
-a. Reactor pressure 30 psig : b. Main steam line pressure of 30 psig c. MSIV sealing system to RPV differential pressure of 8 psid :d. MSIV sealing system flow rate of 770 SCFM - ,
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REACTOR OPERATOR Page 59' i
. QUESTION: 056 (1.00) . !
WHICH ONE (1) of- the following actual or potential exposure i rates is the MINIMUM that require establishing an Exclusion Area? a. 100 mrem /hr b. 1000 mrem /hr c. 10,000 mrem /hr j r d. 100,000 mrem /hr
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.QUESIION:-057 (1.00)
In accordance OP-AP.ZZ-102 (Q) "Use of Operations Department Procedures", ,
- the statement,. " Denotes a recommendation to be adhered to unless conditions warrant deviation" is the definition of WHICH ONE (1) of the following?
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a. Shall I b. -Should a c. May-
'd, Will
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QUESTION:-058~ (1.00) Concerning the reactor manual control . system, if the request signals I from the timer cards are not exactly a'like,'this fault will illuminate
'WHICH ONE (1) of the following lamps on the rod select module?
a. Rod Bypass' b. Rod Not Selected , c. Data Fault
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' :, ; i REACT 0K OPERATOR Page 62 QUESTION: 1059 ~ (1.00) "The reactor had scrammed on a loss of all AC power. The following conditions: exist- - Al1 emergency diesel generators failed to start Reactor _ water level has.just reached -129 and is decreasing .
1 inch / min i
- Reactor pressure is 250 psig and decreasing.10 psi / min
- Reactor _ power is 0% .-
- RCIC_is'in operation- {
- HFil has_ initiated and_ tripped _
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- Drywell' pressure is l_ psig ' ; - ADS has NOT been inhibited With the preceding _ conditions, WHICH ONE (1) of the following j!
describes the response of the ADS system?
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a. The ADS valves will_open in 105 seconds b. The ADS valves will open in 5 minutes _
' c. ,The ADS valves will open-in 6 min 45 sec d. The ADS valves will not open automatically -i o
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Page 63 fffREACTOROPERATOR'. l l-QUESTION: 060 (1.00)'
* .WHICH ONE U p ot'the;following is NOT a core alteration as defined by , Tech' Specs?
a'- LPRM removal
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.b. Removal of the startup source n
'c. Withdrawal. of a SRM detector d. Relocation of'a fuel bundle to the fuel pool [
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_ _ _ _ _ _ _ . . . . _ _ REACTOR OPERATOR Page 64 QUESTION: 061 (1 00) In accordance with OP AB.ZZ-105, " Loss of CRD Regulating Function," WHICH ONE (1) of the following conditions require that the operator manually scram the reactor? _ a. Reactor pressure is 860 psig and two accumulator alarm is present.
_ b. Reactor pressure is 860 psig and charging water header pressure can not be restored within 20 min.
c. Reactor pressure is 900 psig and two accumulator alarm is present.
d. Reactor pressure is 900 psig and charging water header pressure can not be restored within 20 min, l
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. QUESTION: 062 (1.00) WHICH ONE (1) of the following describes the HPCI suppression chamber suction valve (HV-F042) interlock feature?
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ao Automatically opens on a HPCI initiation signal unless a HPCI isolation signal is present b. Automatically opens on a low suppression pool level signal if a HPCI isolation signal does not exist c. Automatically opens on a high suppression pool level signal if the CST suction valve (HV F004) is closed d - Automatically opens on a low CST level if a HPCI isolation signal is not present
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l l QUESTION: 063 (1.00)
Yhe "A" control room ventilation system is operating with the j control room emergency filtration (CREF) in the 0A (Outside Air) mode. A valid control room intake high radiation signal is received. WHICH ONE (1) of the following is the system response? a. No system response is expected because coincidence logic circuitry is not satisfied.
b. Cor. trol area exhaust ventilation will trip. ' c. Both CREF units will start. , d. The CREF units will automatically shift to the recirculation mode.
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REACTOR OPERATOR Page 67 QUESTION: 064 (1.00) WHICH ONE (1) of the following provides the signal for the turbine control valve (TCV) fast closure scram? ao TCV position limit switches b. Rate of TCV position change c. Power to the TCV fast acting solenoids d. ETS oil pressure at the TCV l l l
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QUESTION: 065 (1.00) WHICH ONE (1) of the following loads looses cooling upon a failure [ of the reactor auxiliary cooling system? a. CRD pump oil coolers b. Safety area cooling system room coolers c. Switchgear room coolers j d. Control equipment room coolers
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QUESTION: 066 (1.00) .The reactor operator has manually scrammed the reactor with a steam leak into the main steam line tunnel. The following conditions exist: o Drywell pressure 1.00 psig o Reactor Level - 15 inches .
- Reactor power is 0% - Oxygen concentration in the primary containment is 24% - Nitrogen concentration in the primary containment is 72%
Reactor building ventilation exhaust rad level is
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2.0 EE 03 uti/cc j WHICH ONE (1) of the following conditions caused the isolation of the Hydrogen /0xygen Analyzer System? a. Reactor level b. Drywell pressure ; c. Oxygen concentration d. Rx b1dg vent rad levels
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QUESTION: 067 (1.00) WHICH ONE (1) of the following is the purpose of the voltage regulator of the reactor recirculation flow control system? a. Controls exciter output to produce 70 V/Hz generator volts b. Provides frequency input for V/Hz control c. Provides a linear speed response to a linear speed demand d. Provides recirc motor generator field current
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l QUESTION: 068 (1.00) The plant has experienced a LOCA and you have entered OP E0.ZZ-101 (Reactor / Pressure Vessel Control) under low level. The following conditions exist:
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Reactor Power 0%
- Reactor Pressure 56 psig - Reactor level < -161 inches - Drywell pressure 15.5 psig - Suppression Chamber Pressure 37.5 psig The first overide for Level centrol says: "If Primary containment water level cannot be maintained below 93 ft then...." WHICH ONE (1) of the following is the calculated containment water level? (See Attachment #12)
a. 22.0 feet b. 54.8 feet c. 61.6 feet d. 110.0 feet :
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QUESTION: 069 (1.00) WHICH ONE (1) of the following malfunctions of the Flow L' nits in the Recirculation System will allow rod movement? :
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a. Failure of square root converter downscale i b. Flow differential between loops A and B of 13% c. Flow unit reading 109% d. Flow unit module unplugged
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REACTOR OPERATOR Page 73-QUESTION: 070 (1.00) WHICH ONE (1) of the following automatic functions do NOT occur at a reactor water level of +12.5 inches? a. Reactor scram b. Shutdown cooling isolation c. Permissive to ADS , d. Recirculation pump trip
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REACTOR OPERATOR Page 74 QUESTION: 071 (1.00)
'WHICH ONE (1) of the following conditions will cause the condenser air removal system's mechanical vacuum pump to trip?
a. Main steam line high radiation b. Low main condenser vacuum c. Off gas system high radiation d. High hydrogen concentration in pump discharge
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REACTOR OPERATOR Page 75 QUESTION: 072 (1.00) The plant is operating at 100% power with all three feedwater pumps ! in automatic on the master level controller. WHICH ONE (1) of the following is the response of the plant if the selected level transmitter failed high and no operator action is taken? (Use Attachment #8) a. RFP turbines would lock due loss of level signal input and level would remain about the same, b. RFP turbines reduce speed in response to the high level signal I and level would continue to decrease until the reactor scrams. l l c. Steam and feedwater flow inputs would compensate for the ; level- error signal and level would stabilize at a slightly lower ' level.
d. Level input would automatically transfer to the other level transmitter and level would remain approximately the same. , l
__ REACTOR OPERATOR Page 76 l QUEST!0N:-073 (1.00) The plant is operating at 100% power with all three feedwater pumps
.in automatic on the master level controller. WHICH ONE (1) of the following will be the response of the plant with no operator action if the "A" RFPT control signal fails high? (Use Attachment #8)
a. "A" RFP turbine would lock at its last position b. "B and C" RFP turbines reduce speed in response to high level signal and level will continue to drop until the reactor scrams.
c. "B and C" RFP turbines would compensate for the error signal and
' level would stabilize at a slightly lower level.
d. Control signal for "A" RFP turbine would automatically transfer to the other pumps and level would remain approximately the same.
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REACTOR OPERATOR Page 77 QUESTION: 074'(1.00) The turbine has just tripped at 100% reactor power. The reactor has scrammed and HPCI, RCIC, and reactor feed pumps have all tripped due to a reactor level spike. WHICH ONE (1) of the following conditions is NOT true? a. If the "A" safety auxiliary cooling system (SACS) pump is not running when the turbine tripped it will start at -38 inches Reactor water level.
b. All of the service water pumps will start at -38 inches.
. c. The high reactor water level trip will have to be reset on RCIC for it to start at -38 inches.
d. HPCI will start when wide range reactor level reaches -38 inches
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REACTOR OPERATOR Page 78 QUESTION: 075 (1.00) The plant is operating at 60% power. Three (3) of the four (4) operating running circulating water pumps trip. WHICH ONE (1) of the following is the maximum number of main condenser water boxes allowed to be in service? a. one b. two c. three ; d. four
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REACTOR OPERATOR Page 79 QUESTION: 076 (1.00) WHICH ONE (1) of the following conditions will NOT cause the voltage regulator on the main generator to trip from automatic to manual? a. Loss of voltage feedback signal from the pilot potential transformer b. Excessive volts / hertz c. Generator stabilizer selected to "off" ' d. Exciter field inverse time overcurrent
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QUESTION: 077 (1.00)
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A scram has occurred, the SDV is full and the scram pilot valve
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air header is O psig but there are 21 rods still out and reactor power has stabilized at 12%. In accordance with OPoEO.ZZ-101, "RPV Control" several methods are availible.
WHICH ONE (1) of the following methods should be used to 1 insert the remaining control rods? ,
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a. Individually scram control rod using OP E0.ZZ-303 b. -Isolate and vent the scram air header per OP-EO.ZZ-306, r c. De energize the scram solenoids using OP-E0.ZZ-302.
d. Reset the scram, drain the SDV and. initiate a manual scram.
Defeat RPS interlocks if necessary.
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REACTOR OPERATOR Page 81 QUESTION: 078 (1.00) WHICH ONE (1) of the following is NOT true as it pertains to Hope Creek j tagging procedures? a. Equipment shall not be tagged with both a yellow and a red blocking tag b. Permanent and temporary releases shall be authorized by all persons named on the group tagging request c. Group tagging applies to both red blocking tags and yellow permissive l
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tags d. If there are two or more yellow permissive tags for different l individuals on the same equipment, permission shall be secured from all persons named on the yellow permissive tags before any operation of the equipment is permitted.
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QUESTION: 079 (1.00) WHICH ONE (1) of the following is one of the functions of the Main Steam Line Radiation Monitoring system? a. It monitors the gross gamma radiation from the main steam lines at a location just upstream of the inboard MSIV's, b. It will initiate a PCIS group I isolation, but not a direct scram when the trip level of 3X normal full power background is * reached.
c. It will activate a " Main Steam Line Downscale" annunciator to alert the operator of an equipment malfunction while at power.
d. It will trip the mechanical vacuum pump, if running, and close its suction valves when the trip level of 1.5X normal full power background is reached.
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_ . . . _ REACTOR OPERATOR Page 83 QUESTION: 080 (1.00) WHICH ONE (1) of the following reactor vessel components provide for ' vertical weight support of the peripheral fuel bundles? a. Control rod housing b. Core plate c. -Top guide d. CRD stub tube
REACTOR OPERATOR Page 84 QUE!" ION: 081 (1.00) WHICH ONE (1) of the following loads is supplied electrical power by Class IE 4160 volt bus 10A403? a. TSC water chiller "A" b. Core spray pump C c. RHR pump B d. Station service water pump A
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REACTOR OPERATOR Page 85 QUESTION: 082 (1.00) For one (1) mode of the residual heat removal system, the suction is received from HV-F004A(B), RHR Pump Torus Suction Valve, through the RHR pump then through HV-F047A(B), Heat Exchanger Inlet Valve, then through the RHR heat exchanger, then through HV-F024A(B), Pump Test Return to Torus. WHICH ONE (1) of the following RHR modes does this flow path describe? a. Torus chamber spray b. Containment spray c. Shutdown cooling d. Torus water cooling
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. REACTOR OPERATOR Page 86 f)UESTION: 083 (1.00) In accordance with 10 CFR 26 " Fitness for Duty Program," WHICH ONE (1) of the following is the minimum period to abstain from alcohol proceeding any normal scheduled shif t? a. 3 hours b. 5 hours c. 8 hours d. 12 hours
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REACTOR OPERATOR Page 87 l QUESTION: 084 (1.00) In accordance with SA-AP.ZZ-002(Q) " Station Organization and Operating Practices," WHICH ONE (1) of the following is the normal complement of Fire Brigade for power operations? a. I b. 2 c. 4 d. 6 l-1' l.
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REACTOR OPERATOR Page 88 QUESTION: 085 (1.00) During power operations,1 reactor reci'.culation pumps trips after seven ' Ginutes of operation. In accordance with OP-AP ZZ-109(Q) " Equipment Operation Control," WHICH ONE (1) of the fellowing is the minimum amount of time required before the recirculation pump may be restarted? (Use Attachment #13) a; 15 min b. 30 min c. 45 min d. 60 min - t
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_ REACTOR OPERATOR Page 89 j l QUESTION: 086 (1.00) j During power operations, feedwater heater #6 isolates. The following 1 conditions exist after the isolation:
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- Reactor power 52% increasing slowly 1-l -
Core flow 38%
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85% Rod pattern line
WHICH ONE (1) of the following is.the immediate action in accordance l with OP-AB.ZZ-300 "Rx power oscillations?"
(Use Attachment #9) 1
a. ' Place mode switch in Shutdown ) b. Insert control rods to reduce reactor power to < 40% c. Increase flow to exit this power to flow region d. No immediate actions required for these conditions 1.
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. REACTOR OPERATOR Page 90 i QUESTION: 087- (1.00) J During power operations, with a reduction in forced core flow,
'imediate actions are required if power oscillations are greater than !
10%. In accordance with OP-AB.ZZ-300 " Reactor Power Oscillations," - WHICH ONE (1) of the following is considered normal oscillations for 100% power operations? a. 10 to 8% b. 8 to 6% l c.- 5 to 2% d. 1% or less
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REACTOR OPERATOR Page 91 l l QUESTION: 088 (1.00) WHICH ONE (1) of the following is the 10 CFR 20 limit per calendar quarter for skin of whole body, with quarterly exposure known? a. 1000 mrem b. 1250 mrem c. 5000 mrem
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d. 7500 mrem-
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REACTOR OPERATOR Page 92 i l l LQUEST10N:089 (1.00) During a' loss of instrument air, receiver pressure has decreased to 179 psig. In accordance with OP-AB.ZZ-0131 " Loss of Instrument Air l
.or Service Air", WHICH ONE (1) of the following.should NOT have occured? i a. Emergency instrument air compressor starts I 'b. Standby service air compressor starts ;
c. Service air supply header: isolation valve. (HV 7595) closes
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d. Instrument air dryer isolation valve-(1KB-HV-Il416) opens-f I1
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REACTOR OPERATOR Page 93 QUESTION: 090 (1.00)
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WHICH ONE (1) of the following combinations of Hydrogen, Oxygen, and Nitrogen will result in a Non-flammable mixture?
(Use Attachment #10)
a. H2 - 10%, N2 - 80%, 02 - 10% b. H2 - 30%,_N2 - 67%, 02 - 3% ; c. H2 - 20%, N2 - 70%,.02 - 10% -
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d.- H2 - 15%, N2 - 75%, 02 - 10% '
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REACTOR OPERATOR Page 94 m' r-t-i QUESTION: 091 (1.00) fin' accordance with OP-AP.ZZ-110 "Use and Development of Operating Logs" WHICH ONE (1) of the following describes a Surveillance Log? a. Completed by an operator during the tour of assigned spaces
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'b. Records data that satisfies Tech Spec' requirements
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'" ' c._ Chronologically describes the evolutions'and status _ of the plant d Ed.cidentifies specific evolutions that must be performed on a frequent basis +
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= QUESTION: 092 .(1.00)
In accordance with LP-113-01, " Conduct of Operations", for operational Jcenditions 1,2, or 3 the minimum number of licensed SR0's and RO's combined is? a;-2 ,
b. ' 3 J
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;d. 5 i
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QUESTION: .< 093 (1.00) In accordance with OP-AB.ZZ 121 " Failed Open Safety / Relief Valves," the reactor. mode switch MUST be placed in the shutdown position if an SRV cannot' be closed within a maximum of WHICH ONE (1) of the following times? a. 1: min t b. 2 min t c. 3 min ;
'd. 4 min-l t.' , -!
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QUESTION: 094 (1.00) Ho'e' p Creek is operating at 100% power with a severe thunder storm
. . warning in effect. As the storm intensifies, lighting strikes the i offsite distribution yard causing 500.KV breakers'BS6-5 and BS2-6, and 13.8 KV breakers BS6-7 and BSI-2 to open. Because of this event. the ;- cain turbine. trips and the reactor scrams. WHICH ONE (1).of the following looses power with these given conditions?
a. The'four rod display , b. CRIDS .
. , c. The NSSS computer d. The RSCS display
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' REACTOR OPERATOR Page 98 QUESTIONi 095= (1.00)
AN' inadvertent containment isolation causes drywell pressure to increase.
-In accordance with OP-AB.ZZ-201, "Drywell High Pressure / Loss of Drywell Cooling," WHICH ONE (1) of the following is the required immediate operator action as drywell pressure reaches 1.3 psig? a .- Unisolate chilled water system b. Terminate drywell inerting c. Isolate instrument gas d. Manually scram the reactor
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REACTOR,0PERATOR Page 99 l
, . QUESTION: 0961 (1.00) i ,
OP-E0.ZZ-102A, " Primary Containment Control," step SP/L-7 requires a- . reactor scram if the suppression pool level cannot be maintained above ) 55 inches. WHICH ONE (1) of the following is mitigated by-the reactor scram?- i
a. Unco-cering the drywell to suppression chamber downcomers.
' b. Uncovering the SRV T-quenchers.
Uncovering the HPCI turbine exhaust discharge line,
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d.. Overheating of the suppression pool. ,
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or REACTOR 0PERATOR Page100 , QUESTION: 097 -(1.00): ' 2The' plant is reducing power at a constant rate of 100 MWE per minute , due to a large fuel cladding defact. WHICH ONE (1) of the following
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describes the effect of power reduction on the rate of fission product , release? - a. The release rate increases. exponentially to-the power change
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b. The release rate remains constant with the power change c. The release rate decreases proportionally to the power change l d. The release rate decreases exponentially to the power change i f
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iQUESTION: 098 (1.00)
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WHICH ONE (1) of the following statements describe the boron injection
, initiation temperature as a function of reactor power?
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a. Suppression pool temperature is 110 F at all power levels b '. Suppression pool temperature is 100 F at full power and increases as power decreases c' . Suppression pool temperature is.155 F. at full power and-decreases as power decreases d.. Suppression pool temperature is 155 F at all power levels .;
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Page102
, -QUESTION: 099 (1.00) '
WHICH'ONE (1) of the following states-the reason for an automatic action ,
-at the level 87 f
:
a. Prevent damage to SRVs from potential two phase flow li
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b. . Trip the HPCI. turbine to prevent HPCI turbine blade damage
.c. -Trip the reactor feedpump turbine to prevent overfilling the reactor vessel
.d. Protect the MSIVs from water damage
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t QUESTION: 100 -(1.00) WHICH ONE (1)'of the following automatic isolations of the RWCU system closes ONLY F004 (Outboard inlet isolation)?
a. RPV low level.
b. High Delta flow c. High temp in heat exchanger room
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i jREACTOR'0PERATOR Page 1 ANSWER SHEET Multiple Choice (Circle or.X your choice) If you change your answer, write your selection in the blank.
_ 001 a b c d t 4 002' a b c d t
' ,003 a' b c d a
n .004i a b c d
, : c, . 005 a b c d I ,, [006 a. b c' d
l, ',9'c007 _ a b c d (lj008- a- b c d
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:n'010, a b c d
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f!? ; 013 a b c d l '
:014- a b c d ll 015 a- b c d 016 a- b- c d , ;017 a b c d ll 018 a b .c d-y L 019. a b c d D 020 a b c d >
1, l"' 021 e b c- d E 022 a b c d L 023 a b c d u 024 a b .c d 025 a b c d
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A N'S W E R' SHEET Multiple Choice (Circle or X your choice)
.lf you change your answer, write-your selection in the blank.
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.026 a b c d i
027' a b c d 028 a b -c d
'029- a b c- d
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030 a b c d
:031 a b c d ,032- a b c d-033- a b c. d 034 a b c. d '035 a b c d .036' a b c 'd 037: a b c d 038 a b c d LJ (039- a- b c- d . -040~ a_ b c d '041- a b 'c d 042' a b c d-043 a ~b- c d
' 1044- a- -b c d
}045.- a b c d
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1 : 046 a b c- d "
, 047- a b c d 040' a b c- d : :049 a b c d
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REACTOR OPERATOR- Page 3 ANSWER SHEET-L Multiple Choice (Circle or X your choice)
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If you change your answer, write your selection in the blank.
051 a' b c d
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055 a' b c d 056 a :b c d ! 057 a -b
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058 a- b c d
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059J a b c d , 060 a b. c d 061 a b c d 062 a b- c d t 063: 'a- b c d 064 a b- c- d
;065 a b- c d-
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'0654 a- b- c d
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067: a b c d 068 a' ;b c d
.069 a b c d ?
070. a- b c d 071- a .b. c d 072 a b c- d 073 a b c d
074 a b c d 075 a b c d _ _ _ . _ _ _ _ _ _ _ _ _ _ _ . . _ _
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REACTOR' OPERATOR' Page 4 A N S W-E R SHEET Multiple Choice (Circle or X your choice) If-you. change your answer, write your selection in the blank.
076 a b c d
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077 a b c d 078- a b c d 079- a b c d 080' a b c d r 081 .a b c d.
082;. a b c d
:083 a b c d 084 a b c d ;
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L .085 a b c d !
-086 a~ b c d (.) 087: a b- c d s ' 3088 , a b -c d .089.- a- b -c d I:
L .090 .a b c d [ , i 091- a b c d r h
092' a b c d
;093 a b c d j 094- .a b c d .1 095- a b c d
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'096' a b c -d . 097._ .a- b c. d 098 a b c d 099 a b c d 100 a b c d
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l REACTOR OPERATOR Page 5 .
., 1 ANSWER- SHEET Multiple Choice (Circle or X your choice) - .If'you change your answer, write your selection in the blank.
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REACTOR OPERATOR. Page 1 ,1 ANSWER KEi ! l
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001 a i 002 a
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003- c _; 004~ c i t 005 b: . 006.' d l g-007 a
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008' -d 1 C ; 't 009' 'c l. ; p 010 b n- t 011- c j L 012 b l-013' c , L 0'14 b ., u 015 c-l- '016 .d ,
1: 017- a- 5 m 018. d ; ll 019! c . D 020 c j D 021. c- .. 022. c;
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ANSWER KEY l
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029 c
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030 d 031 a 032 ' d
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033 d-034 c !
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041) 042- ~b
'043 a Lc! " ; b!1 1 045 d n ,
ll :046 d ll 047 b ,
'048 c.
049 d 050 d
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REACTOR' OPERATOR Page 3 ANSWER KEY
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:051 a 052 b 053 b 054. .a C
055 b 056' c 057 b 058 ,d'. 059- d 060 c 061. d-062 d 063 b 064 d-065 a-066- d~ 067- a 068 b"
-069i b 070 d 071 a.
'072 b 073 a 074. c
'075 c
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REACTOR OPERATOR .Page ~4 A N S W'E R KEY
.076. c t
077- d 078 c 079 c 080, b 081- b 082 d i
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084 - d r: 085: c 086 b .
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087 c ! 088 d 089 c , _090 b.
091 b 092 c 093 b
.094' d 095 b-096' a.
097 d 098- a 099 'c . ' 100 d l _ _ _ _ _ _ - _ - - _ _ _ _ _ - . . . _ _ _ _ _ - - - _ - - . - - -
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A N'S W E R -KEY I l
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;. i TEST CROSS REFERENCE Page 1-
. QUESTION VALUE- REFERENCE l
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001. 1.00 8000001-002 1.00 8000002 003 1.00 8000003 I
~004- 1.00 8000005 005 1.00' 8000007 006 1.00- 8000008 007- 1.00 8000009 , =008 1.00 8000010 009 1.00 8000011 :010- 1.00- 8000013 011 l'.00 8000014 '
012 1.00 8000015 013. .l.00 8000016 014 1.00 8000017 015 1.00- 8000018- ,
'016 1.00 8000019 ,
! Ol7L 1.00 =8000020 L 018 1.00 ;8000021 019: 1.00- 18000022
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020 1.00 8000023
'021; 1.00- !8000024 ,
022: ~1.00- 8000025 023 -1.00 8000026 024. 1.00- 8000027-025 1.00 8000028 o 026, 1.00 8000029
.027 '1.00 ? '8000030 028 1.00 8000031
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029 .l '00 , 8000032 ' 030- 1.00' 8000033 ; 031 , l '. 00 ' 7 8000034 032 1.00' 8000035 i 033E - 1. 00. 8000037 .; 034 ~ 1. 00 '- 8000038 035 1.00 8000039 , ' 036 1.00L 8000040 037 1.00 ~ 8000041 038- -1.00 8000042 039 1.00 - 8000043 ' o 040 1.00- 8006044 041 l'.00 80000iG L 042 1.00 8000046 043 1.00- 18000047-i. 044 .1.00 :8000048-045 1.00' 8000049 L 046 1.00~ 8000050 l 047 1.00 8000051 i l
'048 1.00 -8000052 049 1.00 8000053 ; .050 1.00- 8000054 051 'l;00 8000055 052 1.00 8000056 053- 1.00 8000057 054 1.00 8000058 l
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QUESTION VALUE .RE'ERENCE F .
-055 1.00- 8000059- '
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1.00 8000060 057- 1.00 8000061 058 1.00 -8000062
.059 1.00- 8000063 060 1.00 8000064 061 1.00 8000065 ~062 1.00 8000066 '063 1.00- 8000067 064 1.00- 8000068 065 l '. 00 . 8000069 066 1.00 '8000071 4 067 1.00 8000072 068 1.00 8000073-069, 1.00 8000074 070 1.00J 8000075 ,
071' l.00 8000077-072 1.00 8000078 073' l.00 8000079 074- 1.00 8000080 075. 1.00 8000081 n '076 1.00 8000082- ! 077. l.00 8000083- 4
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078 1.00- 8000084-
'079 .l.00 8000085 080 1.00 8000086-081 1.00 8000087 082 1.00 8000088-083 1.00 8000089'
084 -1.00 8000090 085 1.00 8000091 i
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0861 1.00 28000092.
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087 -1.00 '8000093
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L 088 1.00 8000094 E 089: 1.00 8000095 ' i ( :T, '090' -1.00- 8000096 l 091 1.00 8000097 l 1: -092 -1.00 8000098 L' " 093 1.00 8000103 l- 094 1.00 8000104 l 095 1.00 8000105 L ' 096 1.00: 8000106
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097 1.00 8000107 098 1.00= 8000108 099 1.00. 8000109 '
.100- 1.00 8000116 I ......
100.00
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ANSWER:- . 001 (1.00)
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a. [+1'.00] J
.- REFERENCE: '
l" 1.. LP-031 FIG 2 2. 223001G004 (3.6/3.7) t 22300lG004 ..(KA's) i NNSWER: 002 (1.00)- I
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. REFERENCE:
-J d l '. LP-048 UTILITY .
.
L.2. 245000A304-(2.7/2.8)
- :245000A304 ...(KA's)
J t l-! LANSWER:- 003 . (1. 00) . l L c. [+1.00] l i
-
,
REFEREN'CE: ' j ,.
'i l' 1. LP-058 UTILITY 2-. 259001A310 (3.4/3.4)
.
,
259001A310' ..(KA's).
<
9 ANSWER: 004'(1.00) ,
;c. [+1.00]
i 's D
# .__. _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
m
<+- .- ,
.,
I REACTOR OPERATOR ~ Page105
!
REFERENCE: 1.- LP 028-02 Rev 3, pages 24, 32, 21, 29, 34 i 2. . 205000K402 (3.7/3.8) , 205000K402. ..(KA's)
~ ANSWER:. 005 (1.00)
b. [+1.00]
'
REFERENCE: 1. LP 019-09 PG 67,68' 2, 20200lK416 (3.3/3.6)--
-20200lK416 ..(KA's)
ANSWER:- 006 (1.00)
-d.T[+1.00]
REFERENCE: 1. OP-E0-ZZ-307, OP-E0-ZZ-207.
2. 295031K213 (4.1/4.2) i
'
295031K213' ..(KA's)
! ' '
3 ANSWER': - 007.:(1.00)
'
a.[+1.L;
:
'
REFERENCE:
-1. OP-AB.ZZ-130 page 1 2. OP-IO.ZZ-008 page 3 and Attach.-1 ,
3.
' 295016A107.(4.2/4.3) 295016A107 ..(KA's) ,
,
l'
.
v: .
-
y ll?U ; di Vt'?R'EACTOR0PERATOR: Page106 oy; ,, l!:i , '+ .
; , '
L ;- o . - i f;;,[dNSWER:. 008-'(1.00)
, ', "dO[+1'.00] .
o ,
" I REFERENCE:-
s .1, LP 006-07 LO 3 i 3 /2.. 20100lK107.(3.4/3.4)
7 20100lK107 ..(KA's)
,
u ,.
> ' > : ANSWER: 009 (1.00)
If c[+1.00] a.
-c REFERENCE: to * :1._ LP 019-01 pg 28 2; 20200lK407.-(2.8/2.9)
'
hp ; e
* :20200lK407; ' ..(KA's)
F; y,' . ,,a Y lANSWER: 010 (1.00) e,' ..
, - , ' bk [+1.00]
! REFERENCE:-
1. -LP 014-01, pg 21', Fig 8,9 2. 212000K412 (3.9/4.1) , e , 212000K412 ..(KA's)
,
? ANSWER: 011' (1.00)
:
l.".Uc(+1.00] r- u
'
'
.i.
F __ . _ _ _ _ -- - - - -_ , _
_. - __ . . _ _
:. ,
l
'
REACTOR OPERATOR Page107
REFERENCE: 1. .SA-AP.ZZ-024 pg. 32 2. 10CFR20
:3.. 29400lK103 (3.3/3.8) ;
-1 29400lK103 ..(KA's)
'
. ANSWER: 012 (1.00)
ti. (+1.00]
.
1 REFERENCE: i
! 'l . TS 3/4 2-2 2. 215005G005(3.3/4.2)
215005G005 ..(KA's) s i
> ANSWER: :013 (1.00)
c.'[+1.00) l_ LREFERENCE:
-1.:. LP 019-09 pg 22 .
2.- 20100lK410 (3.1/3.0) 201001K410 ..(Kf's) ANSWER: 014' (1.00) b. (+1'.00] a l >
-
. - _ _ _ _ _ _ _ _ - _ _ . .
. REACTOR OPERATOR Page108 REFERENCE:
- -= 1. LP 019 09 pg 32 2. 20100lK503 (2.5/2.6)
k 20100lK503 ..(KA's)
^ ANSWER: 015 (1.00)
c.[+1.00) L: REFERENCE: _ 1. LP 051 Fig 3 " 2. 295005K305(3.8/3.8).
..(KA's) '- 295005K307 ANSWER: 016 (1.00)
.d.[+1.00)
i REFERENCE: 1. LP 051 PG 39 - 2. 241000A107 (3.8/3.7) 241000A107 ..(KA's) _ - 1 ANSWER: 017 (1.00) _ a.[+1.00) _
-
E' f _ W:
- --
: . .
REACTOR OPERATOR Page109 REFERENCE: ! 1. T/S 3/4.4.6 PG 3/4 4-21 2. 295025K105 (4.4/4.7) 295025K105 ..(KA's) ANSWER: 018 (1.00) d.(+1.00) REFERENCE: 1. LP 018-05 pg 34 2. 21500lK401 (2.9/3.0).
21500lK401 ..(KA's) ANSWER: 019 (1.00) c.(+1.00] REFERENCE: 1. LP 060 FIG 6 2, 245000K507 (2.6/2.9).
245000K50/ ..(KA's) ANSWER: 020 (1.00) c.(+1.00] _ _ - _ _ - . ._-_ _____-_--____ ___--_-___--_ _ _ _-__ - -_-_ _
REACTOR OPERATOR Pagel10 REFERENCE:
,
1. OP 10.ZZ 006 PG 17 2. 241000A415 (3.2/3.2) 241000A415 ..(KA's)
;
ANSWER: 02) (1.00) c. [+1.00]
;
REFERENCE: 1. OP-01.ZZ 006 PG 3 2, 241000G005(2.9/3.8) , 241000A415 ..(KA's) ANSWER: 022 (1.00) i
'
c.(+1,00] REFERENCE: e
' '1. LP 005-05 PG 19 2. 201003K407.(3.2/3.2).
201003K407 ..(KA's)
'
l ANSWER: 023 (1.00) b. (+1.00]
, .' REFERENCE:
le LP 054-04 Fig 1 2. 271002G007(3.3/3.4) , 2710000007. ..(KA's)
I w
. .
REACTOR OPERATOR Pagell!
;
i l ANSWER: 024 (1.00) c. [+1.00) I
!
REFERENCE: i 1. LP 023-07 pg 13 l 2. 211000K503 (3.2/3.5)
,
211000K503 ..(KA's)
.
ANSWER: 025 (1.00)
'
b. [+1.0) REFERENCE: 1. LP 014 05 pg 15 fig 12 2. 215003G007 (3.5/3.5) 215003G007 ..(KA's) ANS'JER: 026 (1.00) b.[+1.00) T: 1 - REFERENCE: le LP 005 04 PG 15 2, 201003G007 (3.6/3.6) 201003G007 ..(KA's) . ANSWER: 027 (1.00) b [+1.00]
, - ,
i
.
REACTOR OPERATOR Pagell2 REFERENCE: 1. LP 006 04 Fig 1 2. 20100lK503 (2.7/2.7) 201001K503 ..(KA's) ANSWER: 028 (1.00) d.[+1.00] REFERENCE: 1. LP 030 08 pg 81 2. 217000A202 (3.8/3.7) 217000A202 ..(KA's) ANSWER: 029 (1.00) c.[+1.00) REFERENCE: 1. LP CS2-06 Table 1 2. 262001A301 (3.1/3.2) 262001A301 ..(KA's) ANSWER: 030 (1.00) d.[+1.00)
.
_
- * ..
REACTOR OPERATOR Pagell3 f REFERENCE:- .
'
1. LP 048-07 2. 295005K204 (3.3/3.3)
,
295005K204 ..(KA's) ,
, ANSWER: 031 (1.00)
a. [+1.00] :.
'
REFERENCE: l. LP 017-04 PG 32 2. 215002K401 (3.4/3.5).
215002K401 ..(KA's) ANSWER: 032 (1.00) . , d. [+1.00]
,
REFERENCE: 1. LP 005-04 PG 14
, 2. LP 001-05 PG 29 3. TS B 3/4 1-3 4. 201003G004 (3.5/3.6). -
201003G004 ..(KA's) ANSWER: 033 (1.00) d. [+1.00]
! : , &
: ! . . '
' REACTOR OPERATOR Pagell4 REFERENCE: , 1. LP 001-05 FG 10
'
2. 295009A103 (3.0/3.1).
295009A103 ..(KA's)
;
ANSWER: 034 (1.00) c. [+1.00] t I REFERENCE:
'
1. T/S 3/4 1-2 2. 201003K507 (3.3/3.6)
-
201003K507 ..(KA's)
.
. ANSWER: 035 (1.00) b. [+1.00] REFERENCE: 1. LP 014-05 FIG 7 ,
'2. 215003K401 (3.7/3.7).
' 215003K401 ..(KA's) .
>
ANSWER:. 036 (1.00) c.[+1.00] REFERENCE: 1. LP 068 09 PG 83
~2. 264000K402 (4.0/4.2).
' 26400)K402 ..(KA's)
. .
e __
- - - . - . . . . . . . . . .
REACTOR OPERATOR Pagel15 - ANSWER: 037 (1.00) d [+1.00) L REFERENCE: - 1. LP 022-07 FIG 2 2. 201001K405 (3.8/3.8) 20100lK405 ..(KA's) h ANSWER: 038 (1.00) ' d.[+1.00)
-
REFERENCE: -- 1. LP 044-05 TABLE 1 m 2. 223002K101 (3.8/3.9).
-- 223002K101 ..(KA's)
'
ANSWER: 039 (1.00) -- d.[+1.00) -- REFERENCE: 1. LPL 013-07 PG 30 2. 215004K402(3.4/3.5)
.
215004K402 ..(KA's)
.
ANSWER: 040 (1.00) b.[+1.00) _ E
- 's . ..
REACTOR OPERATOR Pagell6 , k REFERENCE: ,
.
1. LP 013 07 TABLE 2 LO 6
'
2. 215004K401 (3.7/3.7) 215004K401 ..(KA's) ANSWER: 041 (1.00) , b.[+1.00) REFERENCE: f 1. TS 3/4 3.1.4.2 2. 201004G0ll (3.7/4.3). [ 201004G0ll ..(KA's) l ANSWER: 042 (1.00) b.'(+1.00)
.
REFERENCE:. ;
. 1, LP 044 05 TABLE 1 2. 295020K203 (3.1/3.3).
' 295020K203 ..(KA's) ANSWER: 043 (1.00)
!
a. [+1.00] REFERENCE-1.- LP " ALTERNATE LEVEL CONTROL" PAGE-21, LO 4,6 2, 295031K101 (4.6/4.7) 295031K101 ..(KA's)
1
.
, . , < . . -- - - - - - . - . . , . . .-. , - - , . ,- --
- . . . . .
REACTOR OPERATOR Pagell7 i ANSWER: 044 (1.00) be [41.00) REFERENCE: 1. LP FIG 3 051-07 ; 2. 241000A107 (3.8/3.7) 241000A107 ..(KA's) ANSWER: 045 (1.00) d.[+1.00) REFERENCE: 1. LP 048 08 pg 42, LO 3-2. 241000K602 (3.5/3.7) 241000K602 ..(KA's)
:
ANSWER: 046 (1.00) d.[+1.00] REFERENCE: 1. LP 036-05 pg 23 2. 295009G005 (3.7/3.6) 295009G005 ..(KA's) ANSWER: 047 (1.00) b [+1.00) ! _ _ _- , _. .
s , 2 I REACTOR OPERATOR' Pagell8 J REFERENCE: 1. SA-AP.ZZ-002 pg 17 l 2, 294001A103 (2.7/3.7) l 294001A103 ..(KA's) ANSWER: 048 (1.00) c.[+1.00) REFERENCE: 1. LO HC.0P-AP.ZZ 0002 REV 8 PG 20 2. T/S 3/4 0-2 3. 294001A102 (4.2/4.2) 294001A102 ..(KA's) ANSWER: 049 (1.00) d.-[+1.00) REFERENCE: 1. T/S, pg 1-7, 3/4 6-47 2. 29503P.201 (3.9/4.2) 295035K101 ..(KA's) l ' c ANSWER: 050 (l'00)
.
d.(+1.00) l RZFERENCE: l 1. E0P OP-E0.ZZ-101 2. 295037K104 (3.4/3.6) I 295037K104 ..(KA's)
__
.. .
REACTOR OPERATOR Pagel19
.
ANSWER: 051 (1.00) a. [+1.00] REFERENCE: 1. LP 044-05 Table 1 3. 295032K303 (3.8/3.9) 295032K303 ..(KA's) ANSWER: 052 (1.00) b.[+1.00] REFERENCE: 1. LP 044-05 Table 1 2. 295032K303 (3.8/3.9) 295032K303 ..(KA's) ANSWER: 053 (1.00) 1 b.[+1.00]
' REFERENCE:
1. E0P 102A, Primary Containmen't Control 2, 295030G011 (4.3/4.5) 295030G011 ..(KA's)
+
h ANSWER: 054 (1.00) a.-[+1.00]
. . _ . . - ,. .
kCACTOR OPERATOR Pagel20 l
. REFERENCE. -
1. T/S 3/4 2.3. i 2. 295005G004 (2.7/3.7) I l 295005G004 ..(KA's) i j i
;
'
ANSWER: 055 (1.00) l b.[+1.00) REFERENCE: 1. - LP 047-03 pg 11 and Table 1 2, 295032K207 (3.6/3.8) 295032K207 ..(KA's) ANSWER: 056 (1.00) c. [+1.00]
'
REFERENCE: l. SAAP.ZZ-046(Q)pg.4 , 2. 29400lK103 (3.7/3.7) .
<
!
'
29400lK103 ..(KA's) .. ,
i ANSWER: 057 (1.00) b. [+1.00] j. REFERENCE: 1. OP-AP.ZZ-102 "Use of Operations Department Procedures", pg. 3 2.. 294001Alli (3.3/4.3) 294001All! ..(KA's) l.
____-____m - .u.___ ._am_ -_________________m___ __-.-m___.__-
.-.
REACTOR OPERATOR Pagel21 l l l ANSWER: 058 (1.00) d.[+1.00) REFERENCE: 1. LP 00H-007 PG 23 2. 201002A201 (2.7/2.8) 201002A201 ..(KA's)
<
ANSWER: 059 (1.00) d. [+1.00) REFERENCE: 1. LP 029 UTILITY i 2. 295003A103 (4.4/4.4) . , 295003A103 ..(KA's) ANSWER: 060 (1.00) l c.[+1.00) REFERENCE: 1. T/S 1.7 Utility 2. 215004K501 (2.6/2.6) 215004K501 ..(KA's) ANSWER: 061 (1.00) d.[+1.00]
! $ - - ,- -
. - . '
l REACTOR OPERATOR Pagel22 l l l REFERENCE: ) i 1. OP-AB.ZZ-105 UTILITY- j 2, 295022K301 (3.7/3.9) l 295022K301 ..(KA's) ANSWER: 062 (1.00) d.[+1.00] R?.FERENCE: 1. LP 026 09 PG 50 LTILITY 2. 206000K419 (3.7/3.8) , 206000K419 ..(KA's)
;
ANSWER: 063 (1.00) , b'.[+1.00) REFERENCE: 1. LP 096A-01 pg 36 2. 295033A108 (3.6/3.8) 295033A108 ..(KA's) ANSWER: 064 (l'.00) , d. [+1.00)
.
l REFERENCE: 1. LP OE0 UTILITY 2. 212000A210 (3.6/3.8) 212000A210 ..(KA's)
. - _ .
REACTOR OPERATOR Pagel23 I ANSWER: 065 (1.00)
a.[+1.00] 1 l l REFERENCE: l 1. LP 081-08 PG 12 2. 288000K602 (2.5/2.5) 288000K602 ..(KA's) ANSWER: 066 (1.00) d. [+1.00] REFERENCE: 1. LP 034 07 PG 24 2. 295038K302 (3.9/4.2) 295038K302 ..(KA's) \
'
ANSWER: 067 (1.00) a.(+1.00) REFERENCE: i 1. LP 020 UTILITY 2. 20200lK510 (2.8/2.8) 20200lK510 ..(KA's)
.
ANSWER: 068 (1.00) b. [+1.00)
. _ _ . . .- - .
REACTOR OPERATOR Pagel24
' REFERENCE: !
l
'1. E0P 101 CAVTION, E0P 201 CAUTION, 2. 295009G007 (3.4/3.7)
I 295009G007 ..(KA's) ANSWER: 069 (1.00) , b. [+1.00)
-
.
REFERENCE: 1. LP.016-05 PG16 2, 20200lK310 (3.3/3.4)
*-
20200lK310 ..(KA's) ANSWER: 070 (1.00) i ! d.[+1.00] REFERENCE: 1. LP 002 PG 31 UTILITY 2. 295031K209 (3.3/3.4) 295031K?09 ..(KA's) o ANSWER: 071 (1.00) a. [+1.00)
. REFERENCE:
1. LP-053 04 PG 25 2. 245000A203 (3.5/3.6) ; 245000A203 ..(KA's)
- . - .
_ . _ _ _ _ _
' .
REACTOR OPERATOR Pagel25 >
;
I ANSWER: 072 (1.00) l b.(+1.00) I
:
REFERENCE: : 1. LP 059-05 PG27 2. 295009A101 (3.9/3.9) . I
!
295009A101 ..(KA's)
:
ANSWER: 073 (1.00) l a.[+1.00)
,
' REFERENCE: 1. LP 059-05 PG 25 2. 29500lK607 (3.8/3.8)
*
29500lK607 ..(KA's) ANSWER- 074 (1.00) _c.[+1.00) ;
' ,
REFERENCE:- 1. LP-030-07 PG 82 UTILITY I 2, 295005A207 (3.5/3.6) 295005A207 ..(KA's) l t
;
ANSWER: 075 (1.00) c. (+1.00]
. _ .
. _
REACTOR OPERATOR Pagel26
,
REFERENCE: ; 1. HC.0P-AB.ZZ-0125 PG 3 2. 245000G010 (2.8/2.9) 245000G010 ..(KA's) ANSWER: 076 (1.00) c.[+1.00] REFERENCE: 1. LP-061-06 PG 29 2. 245000K407(2.5/2.6) 245000K407 ..(KA's)
>
ANSWER: 077 (1.00) d. (+1.00] REFERENCE: 1. OP-E0.ZZ 101,'RC/Q 16 2. 295015G012 (3.7/4.4) 295015G012 ..(KA's) A ANSWER: 078 (1.00) c.(+1.00) REFERENCE: 1. LP 113-01, LO h.1
'
2. 29400lK102.(3.9/4.5) 29400lK10? ..(KA's)
._ _ _ . .
REACTOR OPERATOR Pagel27
'
ANSWER: 079 (1.00) c.[+1.00) REFERENCE: l 1. LP-221-02 pg 56-61 2. 272000K101 (3.6/3.8) 272000K101 ..(KA's) ANSWER: 080 .(1.00)
.b..[+1.00]
REFERENCE:
.1, LP-001-05 PG 27 2.' 290002G004 (3.2/3.3)
290002G004 ..(KA's) ANSWER: 081 (1.00) b.[+i.00)
. , REFERENCE: ,
L 1.' LP-066-09 TABLE 2 f J2.; 26200lK301 (3.5/3.7)
.
l N 262001K301 ..(KA's) g
-
LANSWER: 082 (1.00)
-
de [+1.00]. l ! l
.. - - _ _ . .
REACTOR OPERATOR Pagel28
;
I REFERENCE: 1. LP-028-07 PG61 2. 219000A102 (3.5/3.5) 219000A102 ..(KA's) ANSWER: 083 (1.00) b. [+1.00) , REFERENCE: 1. 10 CFR 26 2. 294001A103 (2.7/3.7) 294001A103 ..(KA's) ANSWER:- 084 (1.00) d. [+1.00) REFERENCE: 1. SA-AP.ZZ-002(Q) REV 13 PG 22 2. 29400lKil6 (3.5/3.8) 294001K116 ..(KA's) ANSWER: 085 (1.00) c. [+1.00) REFERENCE: 1. OP-AP.ZZ-109(Q) ATTACHMENT 1 PG 3 2. 29400lK107 (3.3/3.6) 29400lK107 ..(KA's)' !
i- ,
. . J REACTOR OPERATOR Pagel29 ANSWER: 086 (1.00)
i b.[+1.00] ' l
,
REFERENCE: 1. 'OP-AB.ZZ-300 PG 3 ' l 2. 29500lG010 (3.8/3.7) l 29500lG010 ..(KA's)
,
ANSWER: 087 (1.00) - c. [+1.00] REFERENCE: ,- 1. OP-AB.ZZ-300 PG 2 l 2, 29500lkl04 (2.5/3.3) ! 29500lK104 ..(KA's) l' l ANSWER:: 088 '(1.00) d. [+1~.00) REFERENCE: i- 1. 10 CFR 20
2. .29400lK103 (3.3/3.8) 29400lK103 ..(KA's) ANSWER:- 089 (1.00) c. (+1.00)
.
I REACTOR OPERATOR Pagel30 REFERENCE: 1. OP-AB.ZZ-0131 PG 1 2. 295019A202 (3.6/3.7) 295019A202 ..(KA's) ANSWER:. 090 (1.00) b.[+1.00] REFERENCE:
'l. LP-103-02 FIG 1 2. 29400lKil5 (3._4/3.8)
29400lKil5 ..(KA's)
' ANSWER: 091 (1.00)
b.[+1.00] ,
REFERENCE: 1. LP 113-01 LO 2 ,- 2. 294001A106 (3.4/3.6) l ! 294001A106- ..(KA's)
,
r ANSWER:. 092 (1.00) c. [+1.00]- REFERENCE: 1. LP-ll3 01 pg 12, LO 12 2.-'.294001A103 (2.7/3.7) 294001A103 ..(KA's)
,
. . . - ._.
. REACTOR OPERATOR Pagel31 i ANSWER: 093 (1.00) b. [+1.00) REFERENCE: 1. OP-AB.ZZ-121(Q) 2. 295026G010 (4.0/3.8) 295026G010 ..(KA's) , ANSWER: 094 (1.00) d.(+1.00) REFERENCE: 1. OP-AB.ZZ-135 PG 2 2. 295003A202 (4.2/4.3) 295003A202 ..(KA's) ANSWER: 095 (1.00) b.[+1.00) REFERENCE: l 11. OP-AB.ZZ-201 PG 1-2. 295020G010 (3.6/3.5) l l l 295020G010 ..(KA's) l ANSWER:- 096 (1.00) a. (+1.00)
_ _ _
. ' !
REACTOR OPERATOR' Pagel32
:
REFERENCE: 1. OP-E0.ZZ-102A PG 19 2. 295030K207 (3.5/3.8) 295030K207 ..(KA's)
.
TANSWER: 097 (1.00) d.[+1.00) REFERENCE: p' 1. LP-101-03 PG 16
! 2. 295017K201 (2.8/3.3)
295017K201 ..(KA's)
- ' ANSWER: 098 (1.00) ,
a. [+1.00) 'i ,' REFERENCE:
-
'
,
1. OP-E0 ZZ-101 RPV CONTROL i 2. ~295013G012 (3.6/4.2) 295013G012 ..(KA's) ; ANSWER: 099.(1.00)' c. [+1.00) REFERENCE:
.l. LP-002-05 PG16 2. 295008K304 (3.3/3.5)
295008K304 ..(KA's) c, . 4 - - - -
_. _ . _ . ._ i
..
REACTOR OPERATOR Pagel33 l
'
ANSWER: 100 (1.00) d. [+1.00] i REFERENCE:
1. LP 021-02, LO 6,7b,4 i 2. 295020K204 (3.1/3.1) l
,
l 295020K204 ..(KA's) l l
,
?
i
:
,
'.
t
' (********** END OF EXAMINATION **********)
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3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING' i
] 3/4.12.1' MONITORING PROGRAM LIMITING' CONDITION FOR OPERATION ,
i 3.12.1. The radiological environmental monitoring program shall be conducted ; as specified in Table 3.12.1-1.
! APPLICABILITY: At all times, i ACTION:- , " a. With the radiological environmental monitoring program not being conducted as specified in Table 3.12.1-1, prepare and subeit to the i Commission, in the Annual Radiological' Environmental Operating Report required by Specification 6.9.1.7 -a description of the reasons for-
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y ; not conducting the program as required and the plans for preventing , a recurrence. . .! b. With'the level of. radioactivity as the result of plant effluents in r an environmental sampling tedium at a specified location exceeding : the reporting levels of Table 3.12.1-2 when averaged over any calendar l quarter, prepare and submit to the Commission within 30 days, p'ursuant
- to Specification 6.9.2, a Special Report that identifies the cause(s) 4 . for exceeding the limit (s) and defines the corrective actions to be .
5-taken to reduce radioactive affluents so that the potential annual dose" to A MEleER OF THE P3LIC is less than the calendar year limits
-
of Specifications 3.11.1.2, 3.11.2.2, and 3.11.2.3.. When-more than , one of the radionuclides in Table 3.12.1-2 are detected in'the : sampling medium, this report shall be submitted if: E concentration (15 concentration (2) L reporting level (3 * reporting level- (2) * ' " > 1 0: j When radionuclides other than those in Table 3.12-2 are detected and.
' are the result of plant effluents..this' report shall be submitted if ,
*
the potential annual dose * to A fSSER OF THE puBLIC from all radio-- L nuclides is equal to or greater then.the calender year limits of specifications 3.11.1.2, 3.11.2.2, and 3.11.2.3. This report is not required if the measured level of radioactivity was not.the result of
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plant effluents; however, in such an event, the condition shall be reported and described in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9.1.6.
.
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c. With milk or fresh leafy vegetable samples unavailable-from one or more of the sample locations required by Table 3.12.1-1, identify-specific locations for obtaining replacement samples and add them to. i the radiological environmental monitoring program within 30 days. '
! "The methodology used to estimate the potential annual dose to a MEMBER OF THE ' ' -
PUBLIC shall be indicated in this report.
"
. L HOPE CREEK 38 12-1 ., . .. . . - - - . , , , ,
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o RADIOLOGICAL ENVIRONMENTAL MONITORING LIMITING CONDITION FOR OPERATION (Continued) ACTION: (Continued) The specific locations from which samples were unavailable may then
.be deleted from the monitoring program. Pursuant to Specifica- ;
tion 6.9.1.8, identify the cause of the unavailability of samples tr.d identify the new location (s) for obtaining replacement samples in the , next Semiannual' Radioactive Effluent Release Report pursuant to Speci-L ! fication 6.9.1.8 and also include in the report a revised figure (s) i and table for the ODCM reflecting the new location (s).
L L d. 'The; provisions of Specification 3.0.3 are not applicable. ! y
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SURVEILLANCE RE0VIREMENTS 4.12.1 The radiological environmental monitoring samples shall be collected pursuant to Table 3.12.1-1 from the specific' locations given in the table and f.igure(s)'in the ODCM, and shall be analyzed pursuant to the requirements of.
Table 3.12.1-1 and the detection capabilities required by Table 4.12.1-1.
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. TABLE 3.12.1-1-5 .
A RADIOLOGICAL ENVIROOMENTAL MONITORING PROGRAM * .
E Number of Representative Exposure Pathway Samples and gy) ' Sampling and Type and Frequency and/or. Sample Sample' Locations Collection Frequency of Analysis , 1. DIRECT RADIATIONI2) Forty-three routine monitoring Quarterly Gamma dose quarterly.
' stations either with two or more dosimeters placed as follows: ' An inner ring of stations','one' ' in'each meteorological sector '
'in the general' area of.the SITE - : BOUNDARY;- * .An outer ring of stations, one M in each land based meteorological .:
J, sector in the 6 .to 8-km range from l the site; and .
'
The balance of the stations to be placed in special - interest areas'such as popula- ! tion centers, nearby residences. l schools, and in one or two i areas to serve as control stations. ;
*
The number, media, frequency, and. location of samples may vary from site to site. . This table presents an - acceptable minimum program for a site'at which each entry is applicable. Local site characteristics must be examined to detemine if pathways not covered by this table may significantly contribute to an individual's dose and sh6uld be included in'the sample program.
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TA8LE 3.12.1-1 (Continued) 5 ~
;g : RADIOLOGICAL ENVIR0fMENTAL MONITORING PROGRAM _
S E'
* Number of . Representative ~ .
Exposure Pathway Samples and Sampling and Type and Frequency and/or Sample Sample Locations (1) . Collection Frequency of Analysis 2. AIRBORNE.
Radiciodine and Samples from 5 locations. Continuous sampler Radiciodine Cannister
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Particulates . operation'with sample 1-131 analysis weekly.
Three samples from close to the - collection weekly, or , SITE 9OUNDARY locations, in more frequently if l different sectors, of a high required by dust Particulate Sampler: . calculated annual average ground- loading. Gross beta radioactivity level D/Q.
, analysis filter changefoli g; 3
$ One sample from the vicinity of a Gamma isotopic analysisI ") -
community having a high calculated of composite (by h annual average groundlevel D/Q. location) quarterly.
a One sample from a control location, '
*
as for example 15-30 km distant and in the least prevalent wind i direction.
3. WATERBORNE a. Surface (5) One sample upstream. Grab sample monthly. Gamma isotopic analysis I4) One sample downstream. monthly. -Composite for One sample crosstream. tritium analysis quarterly.
b. Ground . Samples from one or two sources -Nonthly Gamma isotopic II and tritium only if likely.to be affected(7) . analysis monthly, i
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,% RADIOLOGICAL ENVIR0fMENTAL MONITORING PROGRAM E
ES Number of
* Representative Sampling and Type and Frequency Exposure Pathway Samples and .gg). Collection Frequency of Analysis and/or Sample Sample Locations Composite sample I-131 analysis on each c. Drinking One sample of each of one to .oser 2-week period (6) composite when the dose three the w arest water sup- ' calculated for the consump-when I-131 analysis plies'that could be affected- is performed, monthly- tion of the water is g ater by its discharge. composite otherwise than 1 mres per year. Com-One sample from a control positeforgrossbetaa4- .
gamma isotopic analyses Tocation. monthly. Composite..for tritium analysis quarterly.
Gamma isotopic analysisI #)
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One sample from downstream area Semiannually R d. Sediment semiannually.
- One' sample from upstream' area
.One sample from tross-stream area y . *
4. INGESTION Semimonthly when Gamma isotopic monthlyI4) and a. Nilk Samples from milking animals in I-131 analysis semimonthly three' locations within 5 km animals are on pasture, monthly at when animals are on pasture; distance having the highest dose monthly at other times.
potential. If there are none, other times then, 1 sample from milking animals in each of three areas between 5 to 8 km distant where dosesarecalculatef8}obegreater than 1 mres per yr.
One sample from allking animals at a control location-15-30 km distant.
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RA010 LOGICAL ENVIR0fMENTAL' MONITORING PROGRAM !
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Number of Representative Type aiid Frequency _.
Exposure Pathway Samples and-- Sampling and and/or Sample. Sample Locations (1) Collection Frequency' of Analysis 4. INGESTION (Continued) b. Fish and .One sample of each commercially Sample in season, or Gamma isotopic analysisI *) Inverte- and recreationally important " semiannually if they on edible portions.
i .brates species in. vicinity of plant
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are not seasonal-discharge area.
'One sample of same species in areas not influenced by plant discharge.
' R c. Food One sample of each principal At time of harvest II) Gamma isotopic analysis I43
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Products class of food products from any on edible portion.
M area that is irrigated by water J, in which liquid plant wastes - have been discharged. - I si
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TABLE 3.12.1-1 (Continued) TABLE NOTATIONS (1) Specific parameters of distance and direction sector from the centerline
*
of one reactor, and additional description where pertinent, shall be pro-vided for each and every sample location in Table 3.12.1-1 in a table and l 7 figure (s) in the ODCM. Refer to NUREG-0133, " Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants,' October 1978, I and to Radiological Assessment Branch Technical Position, Revision 1, November 1979. Deviations are permitted from the required sampling schedule if specimens are unobtainable due to hazardous conditions, sea- l sonal unavailability, malfunction of automatic sampling equipment and other legitimate reasons,. If specimens are unobtainable due to sampling equipment malfunction, every effort shall be made to complete corrective action prior to the end of the next sampling period. All deviations from the sampling schedule shall be documented in the Annual Radiological- ' Environmental Operating Report pursuant to Specification 6.9.1.6. It is recognized that, at times, it may not be possible or practicable to con-tinue to obtain samples of the media of choice at the most desired loca-tion or time. In these instances suitable specific alternative media and locations may be chosen for the particular pathway in question and appro-
-priate: substitutions made within 30 days in the Radiological Environmental Monitoring' Program given in the ODCM. Pursuant to Specification 6.14, submit in the next' Sea l annual Radioactive Effluent Release Report documen-tation for a change in the ODCM including a revised figure (s) and table t for the 0DCM reflecting the new location (s) with supporting information ,
identifying the cause of the unavailability of samples for that pathway and justifying the selection of the new location (s) for obtaining samples.
(2)0ne or more~ instruments, such as a pressurized ion chamber, for measuring * l and recording dose rate continuously may be used in place of, or in addi - , tion to, integrating dosimeters. For the purposes of this table, a thermo- ' luminescent. dosimeter (TLD) is considered to be one phosphor; two or more phosphors in a packet are considered!as two or more. dosimeters. Film badges shall not be used as dosimeters for' measuring direct radiation.
The frequency of analysis or readout for TLD systems will depend upon.the L, characteristics of the specific system used and should be selected to obtain optimum dose information with minimal fading.
(3) Airborne particulate sample filters shall be analyzed for gross beta radio-activity 24 hours or more after sampling to allow for radon and thoron , daughter decay. If gross beta activity in air particulate samples is greater than 10 times the yearly mean of control samples,, gamma isotopic analysis shall be performed on the individual samples. !
(4)Gamme isotopic analysis means the identification and quantification of gamma-emitting radionuclides that may be attributable to the effluents from the facility.
(5)The " upstream sample" shall be taken at a distance beyond significant .l influence of the discharge. The " downstream" sample shall be taken in an l m area beyond but near the mixing zone. " Upstream" samples in an estuary ( ') must be taken far enough upstream to be beyond the plant influence. Salt j
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water shall be sampled only when the receiving water is utili:cd for recreational activities.
l HOPE CREEK 3/4 12-7 i
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l l TABLE 3.12.1-1 (Continued) TABLE NOTATION (6)A composite sample is one in which the quantity (aliquot) of liquid sampled is proportional to the quantity of flowing liquid and in which the method' l of sampling employed results in a specimen that is-representative of the I liquid flow. In this program composite sample aliquots shall be collected-at. time intervals that are very short- relative to the compos + ting period in order to assure obtaining a representative sample.
(7) Groundwater samples shall be taken when this source is tapped for drinking ' or irrigation purposes-in areas where the hydraulic gradient or recharge properties are suitable for contamination. .
.(8)The dose shall be calculated for the maximum organ and age group, using i the methodology and parameters in-the ODCM.
- (9)1f harvest occurs more tha,n once a~ year, sampling shall be. performed during u each discrete harvest. If harvest occurs continuously, sampling shall be monthly. Attention shall be paid to including samples of tuberous and root food products.
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" TABLE 4.12.1-1-
~ DETECTION CAPABILITIES'FOR ENVIR0fMENTAL SAMPLE ANALYSIS (1)(2)
h! LOWER LIMIT OF' DETECTION (LLD)(3) E Water Airborne Particulate Fish Milk Food Products Sediment Analysis (pci/1) ' or Gas (pC1/m3 ) (pCi/kg, wet) . (pC1/1) (pCi/kg, wet) (pCi/kg. dry)' gross beta -4 0.01 H-3 3000-Mn-54 15 - 130 !. Fe-59 30 260 l _.
i Co-58,60 15 130 l R! l * Zn-65 30 260 R$
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CD 1-131 1 0.07 1 60
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Cs-134 15 0.05 130 15 60 150 Cs-137 18 . 0.06 - 150 18 80 180 i Ba-La-140 15 . 15
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' l TABLE 4.12.1-1 (Continued)- ' jg TABLE NOTATIONS (1)This list does not mean that only these~nuclides are to be considered.
Other peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9 1.6.
(2) Required detection capabilities for thermoluminescent dosimstars used for environmental measurements shall be in accordance with the recommendations of Regulatory Guide 4.13.
(3)The-LLD is defined, for purposes of these specifications, as the smallest' concentration of radioactive material in a sample that will yield a net count, above system background, that will be detected with 95% probability with only 5% probability of falsely concluding that a blank observation a represents a "real" signal. ;j For a particular measurement system, which may include radiochemical separation: -> _
-4.66 s b '
E - V - 2.22 - Y - exp(-Aat) Where:. ' LLD is the "a priori" lower limit of detection _as defined above, as picocuries per unit mass or volume, s' isLthe standard deviation of the background counting rate or of tNecountingrateofablanksampleasappropriate,ascountsper ' minute,1 , E is the counting efficiency, as counts per disintegration, L V is the sample size in units of mass or volume,
,
2.22 is- the number of disintegrations per niinute per picocurie,
' :Y is the fractional radiochemical yield, when applicable, A is the radioactive dew .onstant for the particular radionuclide ,sec ( 2), and j At-for environmental samples is the elapsed time between sample collection, or end ~of the. sample collection period, and time of counting (sec)
Typical valves of E, V, Y, and at should be used in the calculation.
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HOPE CREEK 3/4 12-11 i
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l' TABLE 4.12.1-1 (Continued)
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I TABLE NOTATIONS It should be recognized that the LLD is defined as an a priori (before the fact) limit representing the capability of a measurement. system and not as an a posteriori'(after the fact) limit for a particular measurement. l Analyses shall be performed in such a manner that the stated LLDs will be achieved ~under routine conditions.. Occasionally background 11uctuations, unavoidable small sample sizes, the presence of interfering nuclides, or other uncontrollable. circumstances may render these LLDs unachievable.
In such cases, the contributing factors shall be identified and described l in the Annual Radiological. Environmental Operating Report pursuant to-Specification 6.9.1.6.
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A VA a m i 0 7 . i 1989 1990 1991 - J A NU A ft Y JULY JANUARY Jtf LY JANUARY JULY 3M T W T F 5 5 M T W T F 5 .5 M T W T F 5 5 M T W T F 5 5 M T W T F 5 5 M T W T F 5 I 2 3 4 5 6 1 9 2 3 4 5 6 1 2 3 a 5 6 7 1 2 3 6 5 I 2 3 4 5 6 e 9 i. n u n u 2 3 0 5 6 e 2 J, , a n n n , , I, it 32 13 1. . , , , ,, it 12 , , , ,, it 12 t, t5 to II 18 59 20 28
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FEBRUARY AUGUST FEBRUARY AUGUST FEBRU4RY AUGUSY t 2 3 4 0 2 3 4 5 t 2 3 1 2 3 4 1 2 1 2 3 9 6 8 9 10 12 4 5 6 7 8 9 19 5 6 7 8 9 60 II 3 4 5 6 7 8 9 4 5 6 7 8 9 to 5 6 7 8 le Il 3 ll 32 13 to 15 le If 88 13 le 85 16 97 IS 19 it 12 93 34 35 16 17 12 13 te 55 16 31 It le 11 32 13 to 15 le It 12 II I4 15 16 IF 59 2e 28 22 23 ft 25 20 ft 22 21 21 25 26 19 39 to tt 22 21 24 19 20 tl 22 21 ft 25 11 in 19 te 21 22 23 19 59 29 ti 12 25 to 26 21 28 27 28 29 3e 31 25 26 27 29 26 21 24 29 30 38 21 25 26 21 ft 25 26 27 29 29 M St M RCH SEPTEMMER MARCH SEPTENBER MARCH SEPTEMBER
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8 2 3 4 8 2 9 2 3 I I I 2 8 2 3 4 5 6 7 5 4 i . 9 3. n 3 4 5 6 . , 4 5 . . . , i. 2 3 4 5 6 i e 2 4 5 6 2 e 9 .
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20 21 22 23 te 22 23 to 25 26 27 2e 19 20 ft 22 21 ft 25 If 18 19 28 ft 22 23 It 19 ft 23 ft 25 26 2r te 29 ft 25 26 ft 29 30 , 26 27 28 29 10 35 23 26 23 28 29 30 25 26 27 28 29 3e 3t 30 38 29 30
$5 A P RIL OCTOBER APRIL OCTOBER APRtt OCT09ER 6 2 3 4 5 6 2 3 4 5 I I 2 3 4 5 6 7 1 2 3 4 5 '6 1 5 2 3 4 5 3 3 2 3 4 5 6 7 0 , , to il 12 t1 le 9 9 19 ll 12 13 to 7 s 9 to it 32 Il 7 8 9 19 18 12 13 e f 8 9 to 11 12 i
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28 29 30 31 29 29 30 27 ft 29 18 38 30 29 1e 38 29 30 MAY NOvtMDER MAY NOVEMBER MAY NOVEMBER 5 2 3 5 2 3 4 8 2 I 2 3 4 5 6 8 2 3 4 8 2 3 4 1 6 9 4 5 6 8 9 5 6 7 8 9 19 Il 3 4 5 6 7 8 9 1 8 9 30 ft 12 13 5 6 2 8 9 to il i 8 le it 82 7 le n 15 06 ti 18 19 26 12 13 to 95 le It it 13 It 15 la II le i9 II 12 il le 85 la If 32 13 to 15 to if te to it 12 IS I4 85 06 25 26 23 19 28 22 25 to 28 22 23 ft 25 26 19 19 26 ft 22 21 ft it to 21 22 23 ft 25 If to 19 28 ft 22 23 ft 22 11 2t ft 23 ft 25 26 27 29 29 30 26
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29 29 39 38 26 25 26 ft 28 29 30 2* 29 10 3t .23 27 28 29 30 27 23 29 30 38 JUNE DECEMDER JUNE DECEMBER JUNE DECEMBER I 2 3 1 2 1 2 I I I 2 3 4 3 6 2 3 5 8 I 8 ' , , 2 3 4 5 6 3 8 2 3 4 5 6 1 0 g , to it 32 33 to 4 5 6 8 9 13 L 4 3 4 5 6 7
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-- - DAT E--- e-07-88
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***PERIDDIC N99 CORE PEReORMANCE LOGE *e
*
'
12 CMWT 3276,
^
6- 7 8 9' is 11 1 2 -3 4 5- PCT PWR 99.5 LOCATION 1.13 1.05 9.91 8.78 9.68 e. 38
' AX1AL REL PWR e.58- 1.18 *1.41, 1.39 1.34 1.26 .
a_n7f-* a.95 _1 22 _.e 95 r" etem
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'
m_ est i' t 90. 8 mi est e e,.gynw CMFCP 1.928 1.18 '1.68 1.3e 1.28 1.96 1.98 e.9e 9.68 1.926
) RING REL PWR CMFLDD i APRM GAF 1. es 1. es 1. et 1. 90 1. 99 1.91 runp.p - t_tsa
' '."'* CMPF 0" 1.646; 4 Y.*i g/f B@'I ff' 6
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e.846re Q.}hl e.761}% "3M i.5 0 ' .761 R[a '9.d46
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1.39 1.48 1.39 1.4e R 189 PMF '1.39 1.49 1 J2E e._s1P CBD
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LOC 21-19- 6 25-14- 6 39-to- 6 19-24- 6 1.82 RW1 34.4e 1.83- 1.85 1.83 1.82 1.83
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l THE NUMPER OF PUNDLES WITH MFLPD GREATER THAN 1. 0 = 4 THE NUMBER OF PUNDLES WITH MAPRAT BREATER THAN 1.0 = s
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L-Containment water 1evel -(in feet) is ' l calculated by: .
-1 -[(suppression chamber' Pressure - Dryvell Pressure) M FT ] + 4.2 TT .
h ' PSI j a) Since the-pressure due only to the height of water is desired, any positive
.
-pressure that exists in-the drywell must L , 'be subtracted.
b) .The 2.3 FT/ PSI conversion factor was
. calculated assuming a 75cr water , temperature, w!.th'a specific volume of . 0161. FT3/L,BM.
L c) The,4.'2.FT facto: was-added because the suppression chambtcr-pressure tap is 4.2
'
FT above the bottor'of suppression
, chamber. Therefore, this height of water .would not be monitorec but must be
considered in the calculation.- -
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OP-A P. L--10 9 (Q ) AlYMtMifMf #3 /
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*
M LARGE MOTOR STARTING 1 CRITERIA , I MOTOR IMFG STARTING RESTRICTIONS COLD MOTOR STARTS HOT MOTOR STARTS Primary Condensate Pump- P.P. 3 consecutlye starts. If. the motor is run at Coastdown to rest is least 30 minutes in an required between starts. hour, it may be started
~ 10 times that hour. If-the motor runs less than 30 minutes in an hour, it may be. started no more than 5 times that hour.
Secondary Condensate Pump P.P. 2 consecutive starts. If the motor is run at Coastdown to - rest is least 30 minutes in an required between starts. hour, it may be started ' 6 times that hour. If the motor runs:less than 30 minutes in an hour, it may be started no more , than 3 times that hour.
Circulating Water Pump P.P. 3 successive starts from- 2 starts'are allowed
-
initial' ambient per hour when motor is at' temperature. 'Coastdown to normal operating rest is required between temperature.
starts. . Note: No more than 6 '! starts allowed in a 24 hr period.
.. s. ; i Rev. 6
~
OP-AP.22-109(0) 1 of 3
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L-
.~ +
-
OP-AP.h -109(Q) 8774(####1 #/3 -
. m mmmmmmer LARGE MOTOR STARTING CRITERI.A MOTOR' ' MFG' STARTING RESTRICTIONS Control Room Chillers Carrier A program timer controls chiller starti'ng and Safety Related Chillers -
Carrier stopping sequence. . Timer provides a minimum of
,
, l 15 minutes interval ^ between starts to allow machine conditions to stabilize.
! Water Chillers Carrier Minimum time between starts is 15 minutes.
Units: A, C'and D Maximum number of starts in a 24 hour period is 6.
~ Note: Upon shutdown following prolonged operation (15 minutes or longer) contro11eks have been designed to permit restart within 3.7 minutes.
Water Chiller Carrier 4 starts per. hour are permissible. Starting Unit: B sequence must be complete prior to additional starts.
Note: Not to exceed 8 starts in 48 hours.
COLD MOTOR STARTS HOT MOTOR. STARTS Station Servit.e West 2 starts allowed from Subsequent starts when Air Compressor ambient-temperature motor has been running are ' permissible at 10 minute intervals. With the motor previously idle, restarts are ' permissible at 30 minute intervals.
. .,
2 o f. 3 Rev. 6
~ OP-AP.22-109(Q)
. - - - ..
.. - . --
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OP-AP. . 1109(Q) < + . n\cMfor9/3J -
.
LARGE MOTOR STARTING-CRITERIA I L
MOTOR MFG' STARTING RESTRICTIONS . COLD' MOTOR. STARTS ' HOT MOTOR STARTS Residual Heat Removal System Pump G.E. 2. successive starts from Service Water Pump G.E. ambient temperature. Motor' can be ' started once Core Spray Pump G.E.- Allow. motor to coast-to at rated temperature.- Motor can be assumed ' to rest'between starts. have returned to rated temperature after 60~
.
minutes of idle time or 30 minutes of running time.
Reactor-Recirculation. Pump Motor
~
Generator G.E. 2 successive starts Motor can be' started once ' from ambient temperature. at rated temperature.
Station Auxiliary Cooling System West- Allow motor to coast to .; Pump Motor can be assumed to i rest between starts. have returned to rated ; temperature after 45 ; minutes of idle time or 15 :! minutes of running time.
Refueling Water Pump Cont 2-starts per hour. I start per hour with Allow motor to coast ~ to motor at normal rest between starts. operational- temperature.
,
OP-AP.22-109(Q) 3 of 3 Rev. 6
. . _ . . . . _ _ ._ . _ . , _ _ ,_ , _ -._ _
. . . _ _ _ - . _ . . . _ _
LREACTOR COOLANT SYSTEM i 3/4.4.4 CHEMISTRY LIMITING CONDITION FOR OPERATION 3.4,4 The chemistry of the reactor coolant system shall be maintained within the limits specified in Table 3.4.4-1.
APPLICABILITY: At all times. i ACTION: '
)
a. In OPERATIONAL CONDITION 1: 1. With the conductivity, chloride concentration or pH exceeding the limit specified in Table 3.4.4-1 for less than 72 hours during one
' continuous time interval and, for conductivity and chloride concen-I; tration,. for less than 336 hours per year, but with the conductivity'-
less than 10 pmho/cm at 25*C and with the chloride concentration less L than 0.5 ppm, this need not be reported to the Commission and the provisions of Specification 3.0.4 are not applicable.
h 2. With the conductivity, chloride concentration or pH exceeding the limit specified in Table 3.4.4-1 for more than 72 hours during one continuous time interval or with the conductivity and chloride l concentration exceeding the limit specified in Table 3.4.4-1 for i L more than 336 hours per year, be in at least STARTUP within the next i
, 6 hours.
L 3. With the conductivity exceeding 10 pmho/cm at 25 C or chloride concentration exceeding 0.5 ppm,.be in at least HOT SHUTDOWN within 12 hours and in COLD SHUTDOWN within the next 24 hours.
b. In OPERATIONAL CONDITION 2 and 3 with the conductivity, chloride concentration or pH exceeding the limit specified in Table 3.4.4-1 for more than 48 hours during one continuous time interval,.be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
c. At all other times: 1. With the: a) Conductivity or pH exceeding the limit specified in Table 3.4.4-1, restore the conductivity and pH to within the limit within 72 hours, or b) Chloride concentration exceeding the limit specified in Table.
3.4.4-1, restore the chloride concentration to within the limit within 24 hours, or perform an engineering evaluation to determine the effects of the out of-limit condition on the structural integrity of the reactor coolant system. Determine that the structural integrity of the
,
reactor coolant system remains acceptable for continued operation prior to proceeding to OPERATIONAL CONDITION 3.
2. The provisions of Specification 3.0.3 are not applicable.
HOPE CREEK 3/4 4-15
.
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--
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,
.
.
,
,
-
. TABLE 3.4.4-1
A REACIOR COOLANT SYSTEM n CHEMISTRY LIMliS E W OPERATIONAL CONDITION CHLORIDES CONDUCTIVilY Qunhos/cm 925'C)' . Pil
' -
1 5 0.2 ppe. $ 2. 9 5.6 $ pil 5 8.6 2 and 3 . 5 0.1 ppe $ 2.0 5.6 $ pit 5 8.6 ~ At all other times , 5 0.5 ppe $ 10.0 5.3 $ pli 5 8. 6 - e. I g t:
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3, '. I, - Attachment 2 ' T. - _ <,-
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?- . . i , Sei 'or Reactor Operator Examination Answer' Key ; t- - t v f. - t, '
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' NRC O' fficial Use Only
-;
I
., .
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; I
.
Nuclear Regulatory Commission 1 Operator Licensing
-
Examination
.
:
This document is removed from i Official Use 0nly category on !" date of examination.
' 1.
NRC Official Use Only ,
'
l.
, < I !
. . - . .. ._ . . _. ___ _
- -
:
..
U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION , REGION 1 - FACILITY: Hope Creek 1
,
REACTOR TYPE: BWR GE4 DATE ADMINISTERED: 90/09/24
-
CANDIDATE: INSTRUCTIONS TO CANDIDATE: - Points for each question are indicated in parentheses after the question. To pass this examination, you must achieve an overall grade of at least 80%. Examination papers will be picked up four and one half (41/2) hours after the examination starts.
' NUMBER TOTAL CANDIDATE'S CANDIDATE'S QUESTIONS POINTS POINTS OVERALL GRADE (%)
............. .......... ............... ...............
' 100 100.00
, l l All work done on this examination is my own. i have neither given ' nor received aid.
, Candidate's Signature
>
-
_ _ _ . _ -_ _ _ . . . _ _ _ _ _ . _ _ _ _ _
- .. <-
i NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination the following rules apply: j 1. Cheating on the examination means an automatic denial of your application and could result in more severe penalties.
2. After the examination has been completed, you must sign the statement on the cover sheet indicating that the work is your own and you have not received or given assistance in completing the examination. This must be done after you complete the examination.
3. Restroom trips are to be limited and only one candidate at a time may leave. You must avoid all contacts with anyt^e outside the examination room to avoid even the appearance or possibility of cheating.
'4. Use black ink or dark pencil only to facilitate legible reproductions.
5. Print your name in the blank providec in the upper right hand corner of-the examination cover sheet.
6. Fill in the date on the cover sheet of the examination (if necessary) 7. You may write your answers on the examination question page or on a separate sheet of paper. USE ONLY THE PAPER PROVIDED AND 00 NOT WRITE ON THE BACK S10E OF THE PAGE.
b. If you write your answers on the examination question page and you need tore space to answer a specific question, use a separate sheet of the paper provided and insert it directly after the specific question. DO NOT WRITE ON THE BACK SIDE OF THE EXAMINATION QUESTION PAGE.
9. Print your name in the upper right-hand co:ner of the first page'of answer ; sheets whether you use the examination question pages or separate sheets of, paper. Initial each of the following answer pages.
10. Before.you turn in your examination, consecutively number each answer sheet,, including any additional pages inserted when writing your answers on the examination question page, 11. If you are using separate sheets, number each answer and skip at least 3 . lines between answers to allow space for grading. 1 12. Write "Last _Page" on the last answer sheet.
! 13. Use abbreviations only if they are commonly used in facility literature.
Avoid using symbols such as < or > signs to avoid a simple transposition
. error resulting in an incorrect enswer. Write it out.
p
!
_ _ . _.
k 14. The point value for each question is indicated in parentheses after the question. The amount of blank space on an examination question page is ' NOT an indication of the depth of answer required.
15. Show all calculations, methods, or assumptions used to obtain an answer.
16. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK. NOTE: partial credit will NOT be given on multiple choice questions.
17. Proportional grading will be applied. Any additional wrong information that is provided may count against you. For exam)1e, if a question is ' worth one point and asks for four responses, eac1 of which is worth 0.25 points, and you give five responses, each of your responses will be worth * 0020 points. If one of your five responses is incorrect, 0.20 will be deducted and your total credit for that question will be 0.80 instead of 1.00. even though you got the four correct answers.
18. If the intent of a question.is unclear, ask questions of the examiner ' only.
L. 19. examination When turning in yourexamination questions, examination, aidsassemble and answerthe completed sheets. examination with In addition, ' turn in all scrap paper.
20. To pass the examination, you must achieve an overall grade of 80% or greater, , l 21. There is a time limit of (41/2) hours for completion of the examination.
(or some other time if less than the full examination is taken.)
22. When you are done and have turned in your examination, leave the examin- ,l ation area as defined by the examiner. If you are found in this area while the examination is still in progress, your license may be denied or revoked.
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QUESTION: 001 (1.00) CRITICAL POWER is defined as: a. The value of the critical power ratio associated with the highest power fuel assembly in the reactor 4: ore.
b. The bundle power which causes critical quality (boiling transition) to exist at some point in the bundle. , c. The power in a fuel rod which causes boiling transition to occur at some point on the fuel rod.
d. The power at which the fuel assemblies achieve boiling transition
'regardless of the quality of the coolant in the core, i
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SENIOR REACTOR OPERATOR Page 5 QUESTION: 002 (1.00)
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The Residual Heat Removal (RHR) System is operating in the Shutdown Cooling (SDC) Mode. WHICH ONE (1) of the following system responses will occur if RPV pressure were to increase to 90 psig? a. Shutdown cooling suction valve (F008) auto closes, RHR pump suction valve (F006 A/B) auto closes and outboard head spray valve (F023) remains open b. All running RHR pumps trip, RHR pump suction valve (F006 A/B) auto closes, outboard head spray valve (F023) remains open c. Shutdown cooling suction valve (F008) auto closes, all running RHR pumps trip, outboard head spray valve (F023) closes d. Shutdown cooling suction valve (F008) auto closes, all running RHR pumps trip, RHR pump suction valve (F006 A/B) auto closes
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QUESTIbN: 003 (1.00) WHICH ONE (1) of the following conditions will cause the automatic ' depressurization system (ADS) valves to open (or remain open)? Assume each event to be unique, a. Core spray pump A and B are operating following a receipt of valid ADS initiating signals.
b. ADS Blowdown is in progress (valves open and initiation signals [ present) and the reactor operator depresses both time reset buttons, c. ADS Blowdown is in progress (valves open) and both ADS solenoid-valves lose electrical power.
d .' Core spray pump B and D are operating following a receipt of valid ADS initiating signals. ,
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SENIOR REACTOR OPERATOR Page 7 , QUESTION: 004 (1.00) WHICH ONE (1) of the fellowing conditions will cause a FULL runback of the recirculation pumps? a. Secondary condensate pump trip with FW flow > 85% b. Primary condensate pump trip with FW flow > 85% c. An RFPT trip (3 running) with RPV level < 30" d. Loss of one circ water pump (3 running) and condenser pressure is 4.5"hg
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_ _ _ _ i SENIOR REACTOR OPERATOR Page 8 i QUESTION: 005 (1.00) Following a required initiation of the Standby Liquid Control system, , you are directed by the " Level / power Control" procedure (OP-E0-ZZ-207) to " Lower RPV water level by terminating and preventing all injection except from CRD and boron injection systems until ....." WHICH ONE (1) of the following conditions are those that allow coolant injection to be re initiated? a. RPV water level reaches TAF, all SRVs remain c,0 sod, drywell-pressure remains below 1.68 psig and SLC tank volume reaches 5% j b. Reactor power drops below 4%, all SRVs remain closed, drywell pressure remains below 1.68 psig, and SLC tank volume reaches 5% c. Reactor power drops below 4%, RPV water level reaches TAF, all SRVs remain closed, and SLC tank volume reaches 5% d. Reactor power drops below 4%, RPV water level reaches TAF, all SRVs remain closed, drywell pressure remains below 1.68 psig .
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__ SENIOR REACTOR OPERATOR Page 9 QUESTION: 006 (1.00) A fire occurs in the Control Room requiring immediate control room evacuation. HPCI- and RCIC auto initiated prior to establishing control at the shutdown panel. At the shutdown panel, all channel transfer switches are turned to "EMER." WHICH ONE (1) of the following
. describes the response of HPCI and RCIC systems?
a. Both HPCI and RCIC continue operating b. HPCI continues operating while RCIC trips c. HPCI trips and RCIC continues operating d. HPCI and RCIC both trip i
SENIOR REACTOR OPERATOR Page 10 QUESTION: 007 (1.00) WHICH ONE (1) of following is the purpose of the positive vent valves of the scram pilot valves in the control rod drive hydraulic system? a. To rapidly interrupt the air supply and provide a vent path for the rapid closure of the scram discharge volume vent and drain valves, b. To direct air pressure to the scram inlet and outlet valves, (. holding the scram valves closed. , I c. To provide a redundant means of bleeding air from the scram inlet / outlet valves. l d. To minimize the air exhaust time from the scram valve air header resulting in quick scram valve opening. '
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QUESTION: 008 (1.00) WHICH ONE (1) of the following signals listed below will cause a' direct recirculation drive motor breaker trip?
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a. Loss of stator cooling (3 5 minute time delay) b. High drywell pressure of 1.68 psig , c. Both M/G ventilation fans trip d. Loss of one (1) cire water pump (3 running) and main condenser pressure of 4.3" Hg
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QUESTION: 009 (1.00) , ' A particular LPRM has a radial X Y coordinate of 24-33, is the second detector from the bottom for this string, and is on switch position 3. On local panel 100608, WHICH ONE (1) of the following describes the identification on the LPRM label? a. 23 24-33 b. 32-24-33
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c. 2D 24-33 d. 3B 24-33 , i
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QUESTION: 010 (1.00) WHICH ONE (1) of the following items will bypass the APRM Upscale Thermal Power trip for the reactor protection system (RPS)? ao Mode switch in Run b. APRM bypass switch in Bypass c. Mode switch in Startup d. IRM range switch on Range 1 :
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. SENIOR REACTOR OPERATOR Page 14 QUESTION: 011 (1.00) .
A tour group is currently visiting Hope Creek during your shift. One of the. group is a 36 year old. This individual is an ex-radiation worker, and has completed an NRC-4 form with a total exposure of 88 REM. This individual has completed the GET and RWT courses and as such his exposure is not limited due to his " visitor" status.
For this individual, WHICH ONE (1) of the following would be the administrative quarterly dose limits? (no extensions) i a. 10 mrem /qtr b. 300 mrem /qtr c. 1000 mrem /qtr d. 2000 mrem /qtr
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QUESTION: 012 (1.00) In calculating the APRM flow biased thermal power-upscale scram trip setpoint, a factor "T" must be determined. WHICH ONE (1) of the ' following states the definition for "T7" a. Average Planar Linear Heat Generation Rate / Linear Heat Generation Rate b. Fraction of Rated. Thermal Power / Core Maximum Fraction of Limiting Power Density
- c. Actual Core Power / Critical Power for Actual Core flow d. Linear Heat Generation Rate / Fraction of Rated Thermal Power
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QUESTION: 013 (1.00) l
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Given the following control rod directional control. valve sequence: (DCV) 123 and (DCV) 121 energize for approximately 2.0 seconds and then de energize, then (DCV) 120 energizes for approximately 4 1.5 seconds. WHICH ONE (1) of the following describes the Rod motion I that resultt? a. Continuous insertion b. Continuous withdrawal c. ~ Single notch insertion d. Single notch withdrawal
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SENIOR REACTOR OPERATOR Page 17 QUESTION: 014 (1.00) WHICH ONE (1) of the following describes the scram discharge volume (SDV) vent and drain valves responce during a reactor scram?
-a. Solenoid operated valves energize to remove control air from the operators of the vent and drain valves, closing them, b. Solenoid o)erated valves de-energize to remove control air from tie operators of the vent and drain volves, closing them, c. Solenoid operated valves energize to supply control air to the operators of the vent and drain valves, closing them, d. Solenoid operatea valves de-energize to supply control . air to the operators of the vent and drain valves, closing them.
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. . SENIOR REACTOR OPERATOR Page 18 QUESTION: 015 (1.00)
WHICH ONE (1) of the following actions will open the main turbine bypass valves after a turbine trip? a. decrease load limit pot b. decrease maximum combined flow pot c. decrease pressure setpoint d. decrease steam throttle pressure
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QUESTION: 016 (1.00) ; The output of both EHC pressure regulators fail high causing a bypass valve to oper,. WHICH ONE (1) of the following EHC controls can be used to close the bypass valve?
a. Bypass jack l b. Load limit c. Pressure set , d. Maximum combined flow limiter
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SENIOR REACTOR OPERATOR Page 20 QUESTION: 017 (1.00) Given the following heat up data: TIME RPV Pressure 7:00 am 233 psig 7:15 am 233 psig 7:30 am 233 psig , 7:45 am 233 psig ' 8:00 am 233 psig 8:15 am 408 psig
.WHICH ONE (1) of the following is the latest time at which :heatup must be secured in order to prevent exceeding the Technical Specification limit for heatup at the current heat up rate?
a. 08:30 am b. 08:45 am c. 09:00 am I d. 09:15 am j
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QUESTION: 018 (1.00)
~00 ring a core traverse by the TIP system with the probe in the core, conditions indicating a Loss of Coolant Accident (LOCA)
occur causing a Group XVII isolation. WHICH ONE (1) of the
' following-statements describes the response of the TIP system?
a. It continues'its sequence for the present core position and then reverses and withdraws into the shield.
b. The ball valve closes regardless of the position of the probe, shearing the probe and isolating the TIP system, c. The TIP system reverses, slow if in the core and fast if i outside the core, withdraws into the shield ! and the ball valve closes. j l d. The TIP system reverses and moves out in fast speed, regardless ) of position, into the shield and the ball valve closes.
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, . QUESTION: 019 (1.00)
A main generator hydrogen leak has occurred. Hydrogen pressure has stabilized at 45 psig. The plant is operating at a 0.95 leading power factor. WHICH ONE (1) of the following is the main generator power output limit based on these conditions? (Use Attachment #2) a. 1000 MW
' b. 1050 MW
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c. 1100 MW d. 1150 MW 1: , l l
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QUESTION: 020 (1.00)
During oower operations with the reactor at 782 MWe, the load ' dispatcher requests that you decrease power to 223 MWe as quickly as possible. WHICH ONE (1) of the following is the minimum time in I cinutes required for this downpower? (Use Attachment #3) a. 2 min b. 9 min ! c. 10 min l d. 20 min
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SENIOR REACTOR OPERATOR Page 24 QUESTION: 021 (1.00) Given the following power history:
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0800-1000 101 %
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1000-1100 98 %
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1100-1300 100.5%-
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1300-1500 99 % WHICH ONE (1) of the following is the maximum percent of power allowed over the next hour without violating Tech Specs? a. 99% , . b. 100% i ( c. 101%- d. 102% l-l
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SENIOR REACTOR OPERATOR Page 25 QUESTION: 022 (1.00) , With the reactor at 100% power, WHICH ONE (1) of the . following conditions could cause a control rod to drift OUT? l a. DCV 120 fails open (settle / direction control valve) b. HCU accamulator depressurizes c. Collet fingers stuck out d. Buffer piston clogged ! l l l l l l l
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QUESTION: 023 (1.00) WHICH ONE (1) of the following selections describes the flow path order for the offgas system? a. Hold up pipe, strip heaters, gas recombiner preheater gas recombiners, gas cooler condenser, glycol cooler condenser, guard bed, adsorber bed, HEPA filters, stack b. Gas recombiner preheater, strip heaters, gas recombiners, gas cooler condenser, hold up pipe, glycal cooler condenser, guard bed, adsorber bed, HEPA liers, stack c. Gas cooler condenser, glycol cooler condenser, gas recombiner preheater, strip heaters, gas recombiners, hold up pipe, guard bed, adsorber bed, HEPA filters, stack d. Guard bed, strip heaters, gas recombiner preheater, gas recombiners, gas cooler condenser, hold up pipe, glycol cooler condenser, adsorber bed, HEPA filter, stack
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SENIOR REACTOR OPERATOR Page 27 l l ' QUESTION: 024 (1.00) WHICH ONE (1) of the following is an assumption for negative reactivity l addition Nsed in the design of the standby liquid control system? t. . 10% shutdown margin b. Todine decay c. Xenon decay
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QUESTION: 025 (1.00) WHICH ONE (1) of the following is the pressure boundary for the IRN detector assembly? a. Shuttle tube .
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b. Dry tube c. Guide tube
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d. Drive tube t'
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ESENIOR. REACTOR OPERATOR Page 29 > QdESTION:' 026' (1.00) WHICH'ONE (1) of the following components of trol rod provide a
, ' path.for reactor water to initiate a scram full out (position 48)
rod when no accumulator pressure is availa61e? I a. Stop piston A0d b. Valve disk-c. Piston tube orifices
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d. Collet assembly
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QUESTION: 027 (1.00) I WHICH ONE (1) of the-following describes the location of the pressure tap inLthe CRDH system for the control- room indication of charging header 1 Cater pressure?~
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a. 'Between the flow control station and the drive water pressure control valve
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b. Between the drive water-filters and the flow control
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station c. Between the pressure control valve and the pressure . equalizing valves p d. Between the CRD pumps and the drive' water filters
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QUESTION: 028 (1.00) WHICH ONE.(1) of the following describes the reset function of the RCIC turbine? a. The:RCIC mechanical overspeed trip can be reset either locally or in the control room, b; The RCIC initiation logic " Reset" pushbutton-resets !
-bothL the manual and automatic-initiation signals.
-c. The RCICLmechanical overspeed trip can be reset from the control room only.
' d. Resetting the trip throttle ' valve .is required for all turbine trips that close the Trip Throttle valve HV-4282 prior to restarting RCIC turbine.
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' SENIOR REACTOR OPERATOR ? - Page 32 . QUESTION: 029 (1.00)
WHICH ONE (1)-of the.following control room annunciators will occur as la direct result of a main generator excessive volt /hz condition? a'. : Power system stabilizer trip b.- Unit protection lockout relay trip , c. . Exciter trip to DC regulator d.- Generator field ground
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. - QUESTION: 030 (1.00)- .WHICH-ONE~ (1) of the following conditions will cause a trip on the main turbine?
a. ETS fluid pressure of 1100 psig at the front standard b. Speed 105% of rated-c. Exhaust hood temperature of 180F
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d .~ Inw shaft oil pump discharge pressure with turbine > 1300 RPM f
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- QUESTION::031- (1.00)
mi120ll CHE (1) of the following conditions allows rod motion? a.. 7 LPRM inputs to the RBM b. Switch in calibrate
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c.- More than one rod selected
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d Failure to null
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QUESTION: 032 (1.00)~ WHICH ONE (1) of the following'is designed to mitigate the consequences
.of a control rod eject' accident?
a. Rod Worth Miniinizer b. Rod Block Monitor c.' Reactor Manual Control , d. Control rod drive support housing
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. QUESTION:.033_ (1.00)-
WHICH ONE (1) of the following is the minimum number of LPRM's needed to consider an APRM channel operable? . t a. 12 ., b. 13 , c. 14 d . 15
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QUESTION: 034 (1.00).
On.a design basis LOCA, certain reactor components ensure two thirds coverage is. maintained. WHICH ONE (1) of the following are NOT one of these components? a. Shroud support ring b. Jet pump-diffuser and mixer c. ' Core shroud d. Core support plate <
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' SENIOR REACTOR OPERATOR Page 38 QUESTION:;035 (.1.00)
During a startup following a refueling ottage, WHICH ONE (1) of the.
-following is the maximum allowable difference between actual rod density and the predicted rod density?
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a. 0.1% Delta K/K b. 0.5% Delta K/K
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c. 1.0% Delta-lK/K d.'10.0% Delta K/K
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QUESTION:. 036.(1.00) ,
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WHICH ONE (1) of the following intermediate range monitor rod blocks is bypassed when the range selector is positioned , to RANGE:I?
.a . 'IRM Hi ,. 'b. IRM Dowiscale c. IRM INOP t
d. Detector Not Full In. ,i L l' i } < o ji
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SENIOR REACTOR OPERATOR Page 40 QUESTION: 037 (1.00)l . e WHICH.ONE (1) of the following conditions will shutdown a diesel
' generator.following an emergency start? ,
a.. Reverse power
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'b. Underfrequency ,
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. c. Bus differential overcurrent -
d. Overvoltage l- .
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WHICH ONE (1) of the following describes the control rod drive 1 hydraulic system response during a scram? a. The scram pilot valve energizes to vent the air off of the scram inlet and outlet valves.
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b. The. scram pilot valve energizes to vent the air off of the scram discharge. volume (SDV) vent and drain valves.
c. The back up scram valves de-energizes to vent the air off of the scram inlet and outlet valves, d. The backup scram valves energize -to vent air off the scram inlet.
and outlet valves.
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_ QUESTION: 039 (1.00)
:- WHICH ONE (1) of the~ following selections will cause a Group;1c MSIV- sealing' system isolation?
a. LReactor b1dg exhaust 1.3 EE -4 uti/cc b. Drywell pressure 1.60 psig
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c. Refuel floor exhaust 1.5 EE -3 uCi d. Reactor vessel level -38 inches
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r QUESTION: 040 (1.00) WHICH ONE (1) of the following statements describes the shorting links that are used in reactor protection system (RPS)? a. Installation of the shorting links activates the SRM Scrams and bypasses the IRM and APRM scrams.
b. Removal of the shorting links activates the SRM scrams in a coincidence of one out-of-two-twice logic scheme, c .1 ' Installation of the shorting links activates the SRM, , IRM and APRM scrams in one-out-of-two-twice logic schemes, d. Removal of the shorting links enables a scram if any single
'SRM, IRM or APRM channel trip. *
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QUESTION:-041 (1.00)
. ' WHICH ONE (1) of the following SRM rod blocks / scrams are bypassed with its associated IRMs on Range 3?
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- a. . SRM Inoperable i-l . b. SRM detector wrong position
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d. SRM detector not-full in I
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SENIOR REACTOR OPER4 TOR' Page 45 QUEST 10N:1042 (1.00) I
~bnilCH ONE (1) of the following ::uld NOT occur as an automatic response to a valid high drywell pressure signal?
a. Instrument gas TIP purge isolation
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b. Recirculation loop B sample 'line isolation ' c. Torus water cleanup / suppression pool supply inboard isolation i d. Containment hydrogen / oxygen analyzer suction isolation
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QUESTION: '043 (1.00). 3
'0P-EO.ZZ-201 " Alternate Level Control" has been entered. RPV level has dropped to -200 inches on the Fuel Zone indication. The procedure requires , *
emergency depressurization. WHICH ONE (1) of the following is the reason for depressurization?. , t
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a'. Maintain peak ~ cladding temperature below 1800F , b. Maintain- total' oxidation of the cladding less than 0.17 of the - total 1 cladding thickness
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c. Maintain'the maximum hydrogen generation less than 0.01 times the hypothetical maximum .
.d. Maintain a coolable geometry for long term cooling l
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' QUESTION: 044 (1.00) /
-GIVENihat:-
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- Load 1 selector ='1067 MWe
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- Load limit - 107 %
. - Max' comb. limit
. 113% /
- Throttle pres. - 947 psig A' pressure reg. - 920'psig
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,9-The .'A' EHC pressure regulator setpoint fails'downscale. (see Attachment #4 is'the resulting bypass valve steam
. flow point #1) WHICH signal.of the EHCONE (1) -of the .following/
logic system? '
'/ i a .- 0% steam flow - / I
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7% steam' flow; -./<- 13% steam flow yP. !
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. ' QUESTION: 045 (1.00)
WHICH ONE (1 of the following states the function of the Intercept valves (IV's ?
'a. Provide backup protection for the main turbine generator if the main turbine stop valves fail b. To direct steam to the fifth point feedwater heaters c. 'To provide temperature control during shell warming '
d. To prevent turbine overspeed upon loss of generator load L' i
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' QUESTION: 046
. (1.00).
- In~accordance.with FSAR and " Containment Hydrogen Recombiner System"
' 'WHICH ONE (1) of the following is the MAXIMUM time allowed before system i < initiation is required after a LOCA event?
a'. Within 8. hours ,
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b. Within-12 hours' ! c. Within 24 hours ; d. Within 34 hours I, c k
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. QUESTIONif047 (1.00)
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Due to vacations ~and required training, you have been asked !
: to work the following schedule next week.
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' Sunday 0800-2000 .' '
Monday 0800 2000 Tuesday 0800-2100 Wednesday 0800-1700
' Thursday '0800-1600 Friday 0800-2100 s Saturday NONE ! ;WHICH'ONE (1) of overtime guidelines, as outlined in SA-AP ZZ-002 ' ." Station Organization and Operating Practices," would be violated.
' Assume times DO NOT include TURNOVER and the plant is operating at
; rated power, a. No more than 16 hours in a 24 hour period .b. No more than 24-hours in a 48 hour period I c.. At least 8 hours rest between work periods d. No more tha'n 72-hours in any 7 day period i
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~ QUESTION:: 048 (1.00)- . ' A Technical Specification Quarterly (92 days) surveillance ,
was performed on the High Pressure Coolant Injection system on
> ' JUNE .10,1989, : SEPT 25,1989 and JAN 1, 1990. WHICH ONE (1) - '
of the following is.the:last MONTH and DAY the next quarterly surveillance can be performed'on? (Use Attachment #7)
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a. Apr'l 1, 1990 b. April 5,1990 c. April 25, 1990 m d. April 30, 1990 I ov i
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l i-QUESTION: 049 (1.00)'
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WHICH ONE (1) of the following is an UNNECESSARY condition for
;Eaintaining secondary; containment integrity?
a.- All secondary containment hatches and blowout panels are closed and sealed b. For double door arrangements, at least one door in each access.to the secondary containment is closed c.. The sealing mechanism associated with each secondary containment penetration is operable d. Maintain the suppression chamber between an indicated r level lof 74.5" and 78.5".
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-QUESTION: 050 (1.00)'
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.DUringanATWSaccident,youareinjectingboronusingtheSBLC
, pumps.J WHICH ONE-(1) of the following tank levels equates to when rCOLD SHUTDOWN BORON WEIGHT" is reached?-
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' a.? l800 gals b.1600 gals
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c. 1400 gals.
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SENIOR REACTOR OPERATOR -Page 54-QUESTION: 051 (1.00) .I WHICH ONE (1) of the following group isolations occur as a result of >a main steam line radiation high-high?
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a. Group 2 Rx recirc water sample system
.b. Group 8 Torus water cleanup system .
c. Group 9 Drywell sumps Ed.. Group.12 Containment atmosphere control ; i ( l l
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_ _ _ _ _ _ _ _ _ _ _ . . . . . . . _ _ . . _ . . i; SENIOR. REACTOR OPERATOR Page 55 m QUEST 10N: 052 (1.00) WHICH ONE (1) of the following group isolation will NOT occur as a result of a drywell pressure high of 1.68 psig? a. Group le MSIV sealing system isolation valves b. Group 2 Rx recire water sample system c.- Group 13 Hydrogen /0xygen analyzer system d. Group 17 Transversing in core probe
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' QUEST'IONi 053 (l'00) / WHICH'ONE (1) of the following conditions require an ' entry: into EOP-102A, " Primary Containment Control?" 'a. Drywell-temp-134 F .
b. _ Suppression pool level 74 inches c. Drywell Pressure 1.2 psig, n .,d. Primary containment Hydrogen 1.68%
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SENIOR REACTOR OPERATOR Page 57 QUESTION: 054 (1.00) ; The function of. the end of cycle RPT is to prevent exceeding WHICH ONE-(1) of the following limits? a. MCPR (minimum critical power ratio) b. MAPLHGR (maximum average planar linear heat generation rate) c. CMFLPD (core maximum fraction of the limiting power density) d. FRTP (fraction of rated thermal power) -
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QUESTION: 055 (!.00) , WHICH ONI. 11 one of the following will prevent initiation of the MSIV sealthy system? a. Reactor pressure 30 psig ; b. Main steam line pressure of 30 psig c. MSIV sealing system to RPV differential pressure of 8 ps'- d. MSIV sealing system flow rate of 770 SCFM
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J QUESTION: 056 (1.00) , WilCH Of.'E (1) of the following actual or potential exposure rates is the MINIMUM that require establishing an Exclusion Area? a.100 mrem /hr b. 1000 mrem /hr c. 10,000 mrem /hr , J. 100,000 mrem /hr
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QUESTION: 057 (1.00) - In accordance OP-AP.ZZ-102 (Q) "Use of Operations Department Procedures", the statement, " Denotes a recommendation to be adhered to unless conditions warrant deviation" is the definition of WHICH ONE (1) of the following? a. Shall b. Should c. May d. Will
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QUESTION: 058 (1.00) The reactor has scrammed on a loss of all AC power. The following conditions exist:
- All emergency diesel generators failed to start o Reactor water level has just reached -129 and is decreasing 1 inch / min
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- Reactor pressure is 250 psig ai,l decreasing 10 psi / min
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- Reactor power is 0%
- RCIC is in operation L - HPCI has initiated and tripped ;
L - Dryvell pressure is 1 psig
- ADS leas NOT been inhibited ,
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l With the preceding conditions, WHICH ONE (1) of the following L describes the response of the ADS system? l l a, 'The ADS valves will open in 105 seconds . l b. The ADS valves will open in 5 minutes i-c. - The ADS valves will open in 6 min 45 sec j l d. The ADS valves will not open automatically
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QUESTION: 059 (1.00) WHICH ONE (1) of the following is NOT a core alteration as defined by Tech Specs? a. LPRM removal b. Removal of the startup source c. Withdrawal of a SRM detector d. Relocation of a fuel bundle to the fuel pool
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SENIOR REACTOR OPERATOR Page 63 i QUESTION: 060 (1.00) In accordance with OP-AB.ZZ-105, " Loss of CRD Regulating function," WHICH ONE (1) of the following conditions require that the operator tanually scram the reactor? p b a. Reactor pressure is 860 psig and two accumulator alarm >j , present.
b. Reactor pressure is 860 psig and charging water header pressure can not be restored within 20 min, p(F c. _ Reactor pressure is 900 psig and two accumulator alarm if present, d. Reactor pressure is 900 psig and charging water header pressure can not be restored within 20 min.
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.,, SENIOR REACTOR OPERATOR. Page 64 QUESTION: 061 (1.00)
WHICH ONE (1) of the following describes the HPCI suppression chamber suction valve (HV-F042) interlock feature? a. Automatically opens on a HPCI initiation signal unless a . HPCI isolation signal is present b. Automatically opens on a low suppression pool level signal if a HPCI isolation signal does not exist , c. Automatically opens on a high suppression pool level signal .L if the CST suction valve (HV-F004) is closed - d. Automatically opens on a low CST level if a HPCI isolation
signal is not present . I
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QUESTION: 062 (1.00) The "A" control room ventilation system is operating with the control room emergency filtration (CREF) in the OA (Outside Air) mode. A valid control room intake high radiation signal is received. WHICH ONE (1) of the following is the system response? a. No system response is expected because coincidence logic circuitry is not satisfied.
b. Control area exhaust ventilation will trip.
c. Both CREF units will start. . d. The CREF units will automatically shift to the recirculation mode.
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I i QUESTION: 063 (1.00) ] WHICH ONE (1) of the following provides the signal for the turbine i control valve (TCV) fast closure scram? 1 a. TCV position limit switches j
b. Rate of TCV position change ,
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c. Power to the TCV fast acting solenoids i , d. ETS oil pressure at the TCV
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SENIOR REACTOR OPERATOR Page 67 i QUESTION: 064 (1.00) WHICH ONE (1) of the following conditions will NOT cause a drywell cooler fan to trip? a. Low flow condition with running fan in SLOW speed b. -129 inches in the reactor vessel I y c. ' Fan motor overcurrent L
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d. . Loss of power signal
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SENIOR REACTOR OPERATOR Page 68 LQUESTION: 065 (1.00)
.The reactor operator has manually scrammed the reactor with a . eam leak into the main steam line tunnel. The following conditions exist: - Drywell pressure 1.00 psig -
Reactor Level - 15 inches
- Reactor power is 0% < - Oxygen concentration in the primary containment is 24% : - Nit. ogen concentration in the primary containment is 72% -
Reactor building ventilation exhaust rad level is
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2.0 EE -03 uCi/cc
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WHICH ONE (1) of the following conditions caused the isolation of the Hydrogen /0xygen Analyzer System? >
a. Reactor level b. Drywell pressure c. Oxygen concentration d. fRx b1dg vent rad levels l r
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SENIOR REACTOR OPERATOR Page 69 i QUESTION: 066-(1.00) The plant has experienced a LOCA and you have entered OP-E0.ZZ-101 (Reactor / Pressure Vessel Control) under low level. The following conditions exist:
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Reactor Power 0%
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Reactor Pressura 56 psig
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Reactor level < -161 inches
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Drywell pressure 15.5 psig
- Suppression Chamber Pressure 37.5 psig - The first overide for Level control says: " If Primary containment water ' level cannot be maintained below 93 ft then...." WHICH ONE (1) of the following is the calculated containment water level? (See Attachment #12)
a. 22.0 feet b. 54.8 feet
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c. 61.6 feet [ . d. 110.0 feet ! R - L, .
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QUESTION: 067 (1.00) WHICH ONE (1) of the following automatic functions do NOT occur at j a reactor water level of 412.5 it. hes? a. Reactor scram , b. Shutdown cooling isolation
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c. Permissive to ADS
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d. Recirculation pump trip
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_. _ SENIOR REACTOR OPERATOR Page 71 QUESTION: 068 (1.00) . WHICH ONE (1) of the following states the reason for maintaining the drywell average temperature below the Tech Spec limit? a. Electrical equipment inside the drywell has been environmentally qualified for below the Tech Spec limit b. The drywell cooling equipment has been designed for operation below the Tech Spec limit. , c. The containment design temperature will not be exceeded during a LOCA.
d. It is consistent with the safety analysis for a transient with small steam leak < 1 inch.
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QUESTION: 069 (1.00)
.The plant is operating at 100% power with all three feedwater pumps in automatic on the master level controller. WHICH ONE (1) of the following is the response of the plant if the selected level transmitter failed high and no operator action is taken? (Use Attachment #8)
a. RFP turbines would lock due loss of level signal input and level would remain about the same.
~ba - RFP turbines reduce speed in response to the high level signal and level would continue to decrease until the reactor scrams, c.. Steam and feedwater flow inputs would compensate for the level error signal and level would stabilize at a slightly lower level.
d. Level input would automatically transfer to the other level transmitter and ?evel would remain approximately the same.
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QUESTION: 070 (1.00) The plant is operating at 100% power with all three feedwater pumps ;
-in automatic on the master level controller. WHICH ONE (1) of the l following will be the response of the plant with no operator action-if the "A" RFPT control signal fails high? -(UseAttachment#8)
a. "A" RFP turbine would lock at its last position
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b. "B and C" RFP turbines reduce speed in response to high level signal and level will continue to drop until the reactor scrams.
c. "8 and C" RFP turbines would compensate for the error signal and level would stabilize at a slightly lower level.
d. Control signal for "A" RFP turbine would automatically transfer to the other pumps and level would remain approximately the same. ,
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.The turbine has just tripped at 100% reactor power. The reacto' -
s scrammed and HPCI, RCIC, and reactor feed pumps have all tr'i:. .ae to a reactor level spike. WHICH ONE (1) of the following cor '. lons is NOT true? - a. If the "A" safety auxiliary cooling system (SACS) pump is not
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running when the turbine tripped it will start at -38 inches Reactor water level.
b. All of the service water pumps will start at -38 inches.
c. The high reactor water level trip will have to be reset on RCIC for it to start at -38 inches.
d. HPCI will start when wide range reactor level reaches -38 inches
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QUESTION: 072 (1.00) , A scram has occurred, the SDV is full and the scram pilot valve air. header is O psig but there are 21 rods still out and reactor power has stabilized at 12%. In accordance with , OP-E0.ZZ 101, "RPV Control" several methods are availible. ' WHICH ONE (1) of the following methods should be used to insert the remaining control rods? a.~ Individually scram control rod using OP-E0.ZZ-303 ) b. Isolate and vent the scram air header per OP-E0.ZZ-306.
c. De-energize the scram solenoids using OP-E0.ZZ-302.
d. Reset the scram, drain the SDV and initiate a manual scram. ! Defeat RPS interlocks if necessary. I l
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; ~0VESTION: 073 (1.00)
WHICH ONE (1) of the following is NOT true as it pertains to Hope Creek tagging procedures? , a. Equipment shall not be tagged with both a yellow and a red blocking I tag b. Permanent and temporary releases shall be authorized by all persons named on the group tagging request c. Group tagging applies to both red blocking tags and yellow permissive tags i i
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d. If there are two or more yellow permissive tags for different individuals on the same equipment, permission shall be secured from all.
persons named on the yellow permissive tags before any operation of the equipment is permitted, s >< t
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, QUESTION: 074 (1.00) In accordance with 10 CFR 26 " Fitness for Duty Program," WHICH ONE (1) of the following is the minimum period to abstain from alcohol proceeding any normal scheduled shift? a. 3 hours b. 5 hours c. 8 hours d. 12 hours
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SENIOR REACTOR OPERATOR Page 78 QUESTION: 075 (1.00) In accordance with SA-AP.ZZ-002(Q) " Station Organization and Operating Practices," WHICH ONE (1) of the following is the normal complement of Fire Brigade for. power operations? o a. I b .- 2 c. 4 d. 6 {
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SENIOR REACTOR OPERATOR Page 79 QUESTION: 076 (1.00) Du' ring power operations, I reactor recirculation pumps trips after seven cinutes of operation. In accordance with OP-AP.Zl-109(Q) " Equipment Operation Control," WHICH ONE (1) of the following is the minimum amount of time required before the recirculation pump may be restarted? (Use Attachment #13) a. 15 min b. 30 min c. 45 min d. 60 min
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' SENIOR REACTOR OPERATOR Page 80 QUESTION: 077 (1.00)
'Ouring power operations, feedwater heater #6 isolates. The following conditions exist after the isolation:
- Reactor power 52% increasing slowly
-- Core flow 38%
- 85% Rod pattern ~1ine
.WHICH ONE (1) of the following is the immediate action in accordance with OP-AB.ZZ-300 "Rx power oscillations?"
(Use Attachment #9)
a. Place mcJe switch in Shutdown b. Insert control rods to reduce reactor power to < 40% c.- Increase flow to exit this power to flow region di No immediate actions required for these conditions
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QUESTION: 078 (1.00) .During power operations, with a reduction in forced core flow, immediate actions are required if power oscillations are greater than 10%.- In accordance with OP AB.ZZ-300 " Reactor Power Oscillations," * WHICH ONE (1) of the following is considered normal oscillations for 100% power operations?
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a. 10 to 8%
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b. 8 to 6%
-c. 5 to 2%
d. 1% or less i
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' SENIOR REACTOR OPERATOR Page 82 QUESTION: 079 (1.00)
"
WHICH ONE (1) of the following is the 10 CFR 20 limit per calendar.
quarter for skin of whole body, with quarterly exposure known? a. 1000 mrem - b. 1250 mrem c. 5000 mrem d. 7500 mrem
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. SENIOR REACTOR OPERATOR Page 83 QUESTION: 080 (1.00) ,
During a loss of instrument air, receiver pressure has decreased to 79 psig. In accordance with OP-AB.ZZ-0131 " Loss of Instrument Air
'or Service Air", WHICH ONE (1) of the following should NOT have occured?
a. Emergency instrument air compressor starts b. Standby service air compressor starts c. Service air supply header 1 solation valve (HV 7595) closes > d.L Instrument air dryer isolation valve (lKB-HV-ll416) opens
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i QUESTION: 081 (1.00) WHICH ONE (1) of the following combinations of Hydrogen, Oxygen, and ; Nitrogen will result in a Non-flammable mixture? l (UseAttachment#10) a. H2 - 10%, N2 - 80%, 02 - 10% l b. H2 - 30%, N2 - 67%, 02 - 3% c. H2 - 20%, N2 - 70%, 02 - 10% l d. H2 - 15%, N2 - 75%, 02 - 10% l l l k l.
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' hUESTION: 082 (1.00) , -In accordance with OP-AP.ZZ-110 "Use and Development of Operating Logs" :WHICH ONE (1) of the following describes a Surveillance Log?
E a.' Completed by an operator during the -tour of assigned spaces
- b.~ Records data that satisfies Tech Spec requirements , ' c. Chronologically describes the evolutions and status o' the plant ,
'F d.' Identifies specific evolutions that must be performed on a frequent ;
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basis
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. . ' QUESTION: 083; (1.00)
i
'In accordance with LP-ll3-01, " Conduct of Operations", for operational
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conditions 1,2, or 3 the minimum number of licensed SRO's and R0's-combined is?-
.a . - 2 ' .
b. 3
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> QUESTION: 084 (1.00)
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In 'accordance with' Technical Specifications, WHICH-Ol:2 (1) of the surveillance frequency notations indicate a frequency of 550 days? _
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-a. R b. - P'
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c. Z
'd. SA .
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~1 ' : QUESTION: 085 (1.00)
The reactor is at 87% power; WHICH ONE (1) of the following Chemistry Reports
'on the . reactor coolant requires that the reactor be placed in hot shutdown e ;within'12 hours? (Use Attachment #14) , , - a. Conductivity .8 umhos/cm, pH 7.3, Chlorides .15 ppm 1"' b. Conductivity 5.3 umhos/cm, pH 8.9, Chlorides .4 ppm c. Conductivity 11.11umhos/cm, pH 6.2, . Chlorides .38 ppm 1- '
d. Conductivity 8.4 umhos/ca, pH 5.9, Chlorides .27 ppm i
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l QUESTION: 086-(1.00).
.WHG, by title, must ultimately give approval to restart after l: . receiving.a startup recommendation from the (SORC) Safety Operations
~ Review Committee?
s-a. L0perations Manager
. b '. NRC Regional Administrator '
l- c. ' Senior Nuclear' Shift Supervisor (SNSS) l
d. General Manager! l l
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i QUESTION:_087-'(1.00)
.In accordance with SA-AP.ZZ-015, " Station Safety Tagging Program,"
WHICH ONE (1) of the following persons is authorized to request a tag
',be removed-from a piece of equipment? '
a. . Senior Nuclear Shift Supervisor 0 b', ' Persons on the Approved Tagging List : , c.- Persons named on the tag d. ~0perations Manager :
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QUESTION: .088 (1.00) In accordance with OP-AB ZZ 121 " Failed Open Safety / Relief Valves," the 1 reactor mode switch MUST be placed in the shutdown position if an SRV
, ;cannot be closed within a maximum of WHICH ONE (1) of the following time w a. 1 min U - b. 2 min i-c. 3 min- ~ d. 4' min , .-
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. QUESTION:t 089 (1.00) i-Hope-Creek is operating at 100% power vith'a severe thunder storm-warning in effect. As the' storm-intens'fies,-lighting strikes the
'offsite distribution yard causing 500_KV breakers BS6-5 and BS2-6, and 113,8 KV breakers BS6-7 and 851-2 to open. Because of this event, the main. turbine trips and the reactor scrams. WHICH ONE (1) of the t ifollowing, looses' power with these given conditions?
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a. The four rod display 3 b; CRIDS ! c. The NSSS computer
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d. - The RSCS display ' a i
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+. -QUESTION: 090 .(1.00)
L AN inadvertent containment isolation causes drywell pressure to increase.
.In-accordance with OP-AB.ZZ-201, "Drywell High Pressure / Loss of Drywell
/ l Cooling," WHICH ONE-(1) of. the following is the required immediate .
operator action as.drywell pressure reaches 1.3 psig?
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a.' 'Unisol' ate chilled water system b. . Terminate drywell inerting- } c. . Isolate instrument gas ) d.1. Manually scram the reactor p
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.QUESTIONi' 091- (1.00) . ;
t' OP-E0.ZZ-102A, " Primary Containment Control," step SP/L-7 requires.a
' reactor: scram if the suppression pool level cannot be maintained above :
55 incher. WHICH ONE-(l) of the following is mitigated by the reactor * scram?
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a'.; Uncovering the drywell-to suppression chamber downcomers.
'b. Uncovering the SRV T-quenchers.
.. c. Uncovering the HPCI_ turbine exhaust discharge line. .
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di Overheating of the suppression. pool. 4 .
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-QOESTION:092-(1.00) 'The. plant is. reducing power at a constant rate of 100 MWE per minute ' due to a large fuel cladding defect. WHICH ONE (1) of the 'following ,
describes the effect of power reduction on the rate of fission product release?' -,
"a. .The release rate increases exponentially to the power change ~ ~
b. LThe release rate-remains constant with the power change y o- .c.- The release rate decreases proportionally to the power change
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d. The release rate decreases exponentially to the power change i l a
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LQUESTION: 093 (1.00)
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WHICH ONE (1)~ of the following statements describe the boron injection.
. initiation temperature as a function'of reactor power? mc a. Suppression pool temperature is 110 F at all power levels ' b.- Suppression pool temperature is 100 F at full power and increases as power decreases c.' Suppression pool temperature in 155 F at full power and-decreases as powerfdecreases
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Ld. l Suppression pool. temperature is 155 F at all power levels i i e
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QUESTION:,094 (1.00) WHICH ONE (1) of the following states'the reason for an automatic action at the ' level 87-
: a. Prevent damage to SRVs from potential two phase flow '
b.. -Trip,the HPCI turbine. to prevent HPCI turbine blade damage
' . c. Trip the reactor .feedpump turbine to prevent overfilling the ;. ,
reactor vessel.
fd. : Protect the MSIVs'from water damage
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! i i SENIOR \ REACTOR' OPERATOR- Page 98 m 4 iQUESTION: 095 (1.00); . Isolation of a:feedwater heater causes n inadvertent reactivity addition' and as a result the SR0 orders a P1 edit to determine if any
-thermal limits nave been violated. 'WHICH ONE (1) of the following statements i describe the thermal limits in violation of Tech Specs? (Use Attachment #11)
L La. 'LHGR is the only thermal limit 1being violated
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b. LMCPR'is the only' thermal 1.imit being violated c. 'APLHGR11s the only thermal limit 'being violated
'ds ;LHGR, MCPR, APLHGR are all .being violated i l
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SENIOR REACTOR OPERATOP Page 99
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QU$STION:096 (1.00)
-! Actions to mitigate reactor power oscillations at high power / low flow conditions are taken to prevent exceeding WHICH _0NE (1) of the following thermal limits? ~
a. MCPR i' b. LHGR i
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c. ~APLHGR d. .PCIOMR.
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:' SENIOR REACTOR OPERATOR Page100 ,
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LQdESTION: 097. (1.00).
[ LIn conjunction with a high reactor pressure WHICH ONE (1) of the following conditions which will cause automatic initiation of the standby
' , : liquid control -system (SLC)?
a. Low reactor level -39 inches and 3.9 min timer timed out
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b. Low reactor level -129 inches and 3.9 min timer timed out
' [ c. _APRMs not downscale and the 3.9 min timer timed out Ld. - APRM' lot downscale and the low reactor level -38 inches e
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EQUESTION: 098 -(1.00)
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A failure to. scram has occurred and the following plant conditions !
," . exist:- ' , -
Reactor power 10%
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Reactor pressureL1020 psig
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Reactor level TAF
- . Drywell pressure 3.2 psig -' Suppression' pool level 77 inches -~
S/U ' level' control is in AUTO
' ' ' ~ Emergency Depressurization (ED) has been directed.'WHICH ONE (1) of the lfollowing systems is used for injection during ED?
_ a. RCIC l:' b; HPCI ' .
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d. Condensate system > ,
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_ - - _ . __ _ _ . . . __ SENIOR REACTOR OPERATOR. ;Page102 i
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QUESTIONi099 (1.00).- The." Minimum zero injection RPV water level" referred to in the bases '
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for OP-E0.ZZ-201, " Alternate Level Control" is WHICH ONE (1) of -the .;
- following?
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, .a. -lowest' RPV water level at _which adequate core cooling _is assured r without steam cooling, b. lowest RPV waterLlevel at which adequate core cooling is assured using steam cooling alone. i L c. : lowest RPV water level using steam cooling which precludes clad temperatures' from exceeding 2200 degrees F. s d. lowest RPV water' level at which initiation of steam cooling
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SENIOR REACTOR OPERATOR Page103
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.QUESTION:.100 (1.00)
-An ATWS. condition exists, Rx power is currently 6% and the main
: turbine is on line.: WHICH ONE (1) of the following statements is N the action for- this plant condition per 0P-E0.ZZ.101, "RPV Control"?
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'a.. Initiate ARI, trip the main turbine if it is on-line, place the mo mode switch'in.S/D .;
b. . Trip the recirc' pumps, trip the main turbine if it is on-line, place the mode switch in S/0 - c.- Initiate ARI, verify recirc runback to minimum, trip the recirc l pumps.
' d. . Trip the recirc pumps, initiate ARI, trip the main turbine if it is.on-line
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b SENIOR REACTOR OPERATOR Page 1 , ANSWER SHEET Multiple Choice (Circle.or X your choice) If you change your answer, write your selection in the blank.
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001 a b' c d
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002 a .b c d 7 003 a b. c d 1 004- .a b' c d
.005- a b c d :l
006; a' b c d l 10071 a- -b -- c d
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l 009 .a b c. d- I o l _ 010; .a .b- c d-
-011- a b c: d !
: 012.- a- b c d-g 013 .a- b c. d _
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014' a b- .c. d i E o 015 a: b c d o , l 016-~ a .b' c d L3 a 017 - a b- c' d Lh018- a: 1 c d ! h' 019.: 'a, b c d , 020- a b c d
.-021; a b c d
. 022: a b c d~
h- 023' a b c- d 024 a b c d 025 a b c d a
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SENIOR REACTOR OPERATOR: Page 2 ANSWER SHEET
' '
Multiple . Choice (Circle or X your choice)' If you change your answer, write' your selection'in the blank.
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.026 a. b c d l 027- a b c. d - 028.- a. b c d c- ' (029l a. b d
' , 030. a b c d t .031 a b: c d .0321 a- b c d '
033' a b c- d-034. a b c: d l r
'035 a- b- c d 1036 a b c d 1 ' < 037 a b c. d , , !038: .a. b c. d l i
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'040 a b< .c d
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% SENIOR REACTOR OPERATOR : Page 3 y
j. ANSWER S li E E T Multiple Choice' -(Circle or X your' choice)
, ' 'If you change your answer, write your selection in the blank.
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' 051 a b' c d
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'O 2: a b c d 053' a' b. c d
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054' .a1 b c- d 055 a b- c d s.056 a. b c d
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059i :a b c d. , L -- 060 a b c d 1 e . , 061" a b' -c d i
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" 066 - .a b- c d ._ j 0671 .ai b c d i u'(068 v. v a. b c d .i. 069 - !a- .b- c d i.' '070 -: a . .b c d W "
L ' '071J a. b c d' J 072 a b c d 073 a b c d
'074' a b c 'd 075' a .b c d j ,
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A N S W E R.- SHEET
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- Multiple Choice - (Circle or X your choice)
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.If you change your answer, write your selection in the blank, i
'076 a b' c d 077 a b c d
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078 .a b c d
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079 a b c d
'080: a b c d- > a- .b. .c ' '081:. -
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082-- .a -b c d 083 ~'a - b c- d
,084, a, b' c d 085: a b c. d 086~ a. -b 'c d ,
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090- ,a . b' - c- -d l > : 091 ' .a .. b - c- d 092: .a b c d
:093 .a b' c d
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094 a b c d 095 ' <a b- c d
' !096' :a- b c d-
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p .098- a b c d ! I: ! 099 a b c d
) '100 a b c d l l I o
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SEN[0RREACTOROPERATOR: Page 1 .t-A N'S W E d - KEY
.; .,. -L001 b 002' c.
- 003- =d , 004 b
.005 d-006 a
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y ( l .- 007 d-008' c 009. d' . 0102 b
'011- c- -012- b
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>015 c L .
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017 >a 018: 'd 019'- c-020 'c 021- c 022~ c 023 b 024 c
'025 'b' ,
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A'N S W E R- KEY
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,026 'b-027 b 3028 d
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! .' .029. , c L .- D' 1030 d !! n L, 031 a-1,
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:. 033 - c 034?: -- d - , "035 '
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036- b n '037- c L' L 038 'd 1039 d' ll-0401'S 'd .
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' A'N S'W E R. K'E Y
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051 a 052: b 053' b t 054 a 055 b- .I 056 c
-057; b-058 d i-059- -c >
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12 070 a 071! c I 072' d 1073 c 074 'b' 075 d
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i ,' SENf0R REACTOR OPERATOR. Page. 4 ANSWER K E'Y
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, . . -076: c-9 077 b 7.:: 078: c 1079 d 080 c- - 081 b ' '-082 b-083. c ' -
084 a
; 085' .c 086- d 1 087- c-088 b-089 'd '- + ' >091' a 092'- , d -' . 093' .a ' 094:' c 095- d- - 096 .a .
097 c
.098 a 099 b.
, :100 c
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.e 4: , 'y: ., , SENIOR REACTOR OPERATOR -
Page 5
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EA.N S'W E R~ KEY
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TEST CROSS Per:RENCE Page 1 QUESTION VA.95 REFERENCE __ 001. 1.00 8000004 002 1.00 8000005
? 003 1.00 8000006 004 1.00 8000007 ^
005' 1.00- 8000008 006 1.00 0000009 007 1.00 8000010 008 1.00 8000011 " 009 1.00 8000012 010 1.00 8000013
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011 1.00 8000014 012 -1.00 8000015 -
013 1.00 8000016 014- 1.00 8000017 - 015 1.00- 8000018 -- 016 1.00 8000019 1.00 8000020 - 017
'018 1.00 8000021
_ 019 1.00 8000022 e 020 1.00 8000023 - 021 1.00 8000024 022 1.00 8000025 023 1.00 8000026 024 1.00 8000027 - 025. 1.00 8000028 - -- 026~ 1.00 8000029 " 027 1.00 8000030 028 1.00 8000031
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029 1.00= 8000032 1030 1.00 8000033-
-031 1.00 8000034 032 1.00 8000035 ;-
033- 1.00 8000036 034 1.00 8000037 035 1.00 8000038
:- 036 1.00- 8000039 037 1.00 8000040 038 1.00 8000041
_ 1039- 1.00- 8000042 040 1.00' 8000043 041 1.00 8000044 042 1.00 8000046 043 1.00 8000047 044. 1.00 8000048
~ 04 5, 1.00 8000049 >
046 1.00 8000050
? .047. 1.00 8000051 048 1.00 8000052 049 -1.00 8000053 050- 1.00 8000054 051 1.00 8000055-052 1.00 8000056 053 1.00- 8000057 054- 1.00- -
8000058 _
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! ' , -TEST CROSS REFERENCE Page 2 QUESTION VALUE REFERENCE ;
055 1.00 8000059 , 056 1.00 8000060 057 1.00 8000061 058 1.00 8000063 059 1.00 8000064 060 1.00 8000065 061 1.00 8000066 062 1.00 8000067 063 1.00 8000068 064 1.00 8000070-065 1.00 8000071-066 1.00 8000073 o- 067 1.00 8000075 068' 1.00 8000076 069- 1.00 8000078 i.
' 070 1.00 8000079 071 1.00- 8000080 072 1.00- 8000083 073 1.00 8000084 074 1.00 8000089 075 1.00 8000090 1 076 1.00 8000091 L 077 1.00 8000092
'078 1.00 8000093 079 1.00 .8000094 080 1.00 8000095 '081 1.00 8000096 '
l' 082 1.00 8000097 l 083 1.00 8000098 ' l- 084 1.00 8000099 l 085 1.00L 8000100 l :086 1.00 8000101 L 087 1.00 8000102 4 ! 088- 1.00 8000103 089. 1.00 8000104 .
.090- '1.00 8000105 *
091. 1.00. 8000106 092 1.00 8000107 093 1.00: 8000108 L 094- 1.00 8000109 095 1.00 8000110 096 .].00: 8000111 " 097. 1.00 8000112: lL ~098 1.00 8000113 099 1.00 8000114 100 1.00 8000115
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EEN!OR REACTOR OPERATOR Page104
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ANSWER: -001 (1.00) b[+1.00) REFERENCE: . 1. Tech Specs 1-2 2. '20100lG005 (3.3/3.9) 20100lG005 ..(KA's) e
= ANSWER: 002- (1.00)
c.[+1.00) REFERENCE: , 1. LP 028 02 Rev 3, pages 24, 32, 21, 29, 34 2. 205000K402 (3.7/3.8).
205000K402 ..(KA's) i ANSWER:> 003 (1.00)
'd.[+1.00]
REFERENCE: L 1. LP 029 02 pages 20/21, 17, 16, 10 L , 2. 218000K403 (3.8/4.0) 218000K403 ..(KA's) l- ANSWER: 004 gl.00)
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REFERENCL: 1. LP 019 09 PG G7,68 2. 20200lK416 (3.3/3.6) 20200lK413 ..(KA's)-
' ANSWER: 005 (1.00)
d [+1.00]
,
REFERENCE: 1. -0P E0-ZZ-307, OP-EO ZZ-207 2. 295031K213 (4.1/4.2) 295031K213 ..(KA's) ANSWER: 006 (1.00) a. (+1.00)
.
REFERENCE: ; l ! 1. OP-AB.ZZ 130 page 1 ,
' 2. OP-10.ZZ-008 page 3 and Attach, 1 ! '
3. 295016A107 (4.2/4.3)
.295016A107 ..(KA's) [
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. I REFERENCE:i l' .l. LP 006 07 LO 3 2.- 20100lK107 (3.4/3.4)
20100lK107- ..(KA's) ,
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I SENIOR REACTOR OPERATOR Page106 , ANSWER:. 008 (1.00) , c(+1.00)
' REFERENCE: 1 1. _LP 019-01 pg 28 2. 20200lK407 (2.8/2.9)
l 20200lK407 ..(KA's)
. ANSWER: 009 (1.00) I d[+1.00]
: REFERENCE: ,
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l'. 'LP 015 01 pg 20 2. 215005A303 (3.3/3.3) 215005A303 ..(KA's) ANSWER:- 010 (1.00) I b..(+1.00] ; i k i REFERENCE:
L -l. LP 014-01, pg'21,' Fig 8,9 t l '- 2. 212000K412 (3.9/4.1) I : r-
212000K412 ..(KA's) 1 - o '
:: ANSWER: 011 .(1.00)
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SEN!OR REACTOR OPERATOR Page107 REFERENCE:. l.. SA AP.ZZ-024 pg. 32 2. 10CFR20 3. 294001K103 (3.3/3.8) 29400lK103 ..(KA's) ANSWER: 012 (1.00) b. (+1.00) REFERENCE: 1. TS 3/4 2 2 2. 215005G005(3.3/4.2) 215005G005 ..(KA's) ANSWER: 013 (1.00) c.[+1.00)
REFERENCE: 1. IP 019-09.pg 22 2. 20100lK410 (3.1/3.0)- 20100lK410 ..(KA's) LANSWER: 014 (1.00)
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l REFERENCE: ! l 1. LP 019 09 pg 32 I 2. 20100lK503 (2.5/2.6) . I 201001K503 ..(KA's) i ANSWER: 015 (1.00)
-c. [+1.00]
REFERENCE:-
,- 1.s LP 051 Fig 3 . ! '
2. 295005K305-(3.8/3.8). , 295005K307- ..(KA's) . ANSWER: 016 (1.00) , d.[+1.00) > REFERENCE:
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1. LP 051 PG 39 2. 241000A107 (3.8/3.7) l 241000A107 ..(KA's)
' ANSWER: 017--(1.00)
a.'[+1.00] l '
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SENIOR REACTOR OPERATOR Page109 REFERENCE: 1. T/S 3/4.4.6 PG 3/4 4-21 2. 295025K105(4.4/4.7)
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295025K105 ..(KA's) ANSWER: 018 (1.00) d.[+1.00) I
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REFERENCE:
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1. LP 018-05 pg 34- l 2. 215001K401 (2.9/3.0).
21500lK401 ..(KA's) ANSWER: 019 (1.00) c.'(+1.00]
. REFERENCE:'
l 1. LP 060 FIG 6 2.- 245000K507 (2.6/2.9).
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REFERENCE: l 1. OP-10.ZZ-006 PG 17 i 2. 241000A415 (3.2/3.2) l
1 241000A415 ..(KA's) l l
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' ANSWER: 021 (1.00) ,
'
c. [+1.00) REFERENCE: , 1. OP-01.ZZ-006 PG 3 2, 241000G005 (2.9/3.8)
'241000A415 ..(KA's)
ANSWER: 022 (1.00) -
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.c. [+1.00]
;
. REFERENCE:.
l.
' 1. LP 005-05'PG 19 2. 201003K407 (3.2/3.2).
E201003K407' ..(KA's)-
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ANSWER: -023 (1.00)
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! Lb.:(+1.00) l REFERENCE:L
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l. 'LP 054-04 Fig 1 Ll 2.- 271002G007 (3.3/3.4) l-
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1 ' l 271000G007 ..(KA's) l l 1:
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_- . . - - _ _ _ - . . - . . . . _ . . .,
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SEN!5RREAbTOROPERATOR 'agell! ANSWER: 024 (1.00)
.c.[+1.00) .
REFERENCE:
-1. LP 023 07 pg 13 2. 211000K503 (3.2/3.5)
211000K503 ..(KA's)
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ANSWER: 025 (1.00) b.-[+1.0) REFERENCE: 1. LP 014 05 pg 15 fig 12 2. 215003G007 (3.5/3.5)
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215003G007 ..(KA's) ANSWER:, 026=(1.00)
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b.[+1.00) REFERENCE:
'l._ LP 005 04 PG 15 2. 201003G007 (3.6/3.6)
201003G007_ ..(KA's) ] ' ANSWER: 027. (1.00) b.[+1.00):
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k
' - SENIORREkCTOROPERATOR Pagella
REFERENCE: 1. LP 006-04 Fig 1 2. 20100lK503 (2.7/2.7) 20100lK503 ..(KA's) ANSWER: 028 (1.00) d.[+1.00] REFERENCE:- 1. LP 030-08 pg 81 2. 217000A202 (3.8/3.7) 217000A202 ..(KA's)
- ANSWER: 029 (1.00)
c. [+1.00]-
,
REFERENCE: 1.. LP 062-06 Table 1 2. 262001A301-(3.1/3.2) 262001A301. ..(KA's) ( MSWER: 030 (1.00) <
,
d-[+1.00)'
: , _ _ - . _ . . _ . . . _ _ _ _ _ _ . . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ .
- _ __ ._ - . .o.
SENIOR REACTOR OPERATOR Pagell3 REFERENCE: 1. LP 048-07 2. 295005K204 (3.3/3.3) 295005K204 ..(KA's) ANSWER: 031 (1.00) a.[+1.00] REFERENCE: 1. LP 017-04 PG 32 '
.
2. 215002K401 (3.4/3.5).
215002K401 ..(KA's) . ANSWER: 032 (1.00) , d. [+1.00)- REFERENCE: , l .- LP 005-04 PG 14 2. LP 001-05 PG 29 ,
~3. ,TS B 3/4-1-3-4. 201003G004 (3.5/3.6).
. 201003G004 ..(KA's)
,
ANSWER: 033 (1.00) c. [+1.00]
. . . ,q
. . - . .. ,
I'
' -. / ' ,
SENIOR REACTOR OPERATOR Pagell4
'
,
'
. REFERENCE:
.
I '. LP 015, LO 3b 2. 215005K116'(3.3/3.4) 215005K116 ..(XA's) ANSWER: 034-(1.00) d.'(+1.00] REFERENCE: 1. LP 001-05 FG 10 2. 295009A103 (3.0/3.1).
295009A103 ..(KA's) ANSWER: 035 (1.00)- c. (+1.00] REFERENCE:
li 1. :T/S 0/4 1-2 2. 201003K507 (3.3/3.6)- 201003K507- ..(KA's)
'
<
ANSWER:- 036 (1.00) b. (+1.00]- REFERENCE: 1. LP 014-05 FIG 7 2. 215003K401 (3.7/3.7). , 215003K401 ..(KA's)
.,
_, - . _ _ . . __ . __. _ SENIOR REACTOR OPERATOR' Pagell5
,
ANSWER: 037 (1.00) >
.
c. [+1.00] REFERENCE: 1. LP 068-09 PG 83 2. 264000K402 (4.0/4.2).. , 264000K402 ..(KA's) ANSWER: 038 (1.00) d.[+1.00) REFERENCE:
{
1. LP 022-07 FIG 2 2. '20100lK405 (3.8/3.8) ' 201001K405 :..(KA's) ., ANSWER:= 039 (1.00)- L' ' d.[+1.00).
REFERENCE:
' -1. LP 044-05 TABLE 1 .
I q 2. 223002K101 (3.8/3.9).
223002K101- ..(KA's) L ANSWER: 040 (1.00) l
>
d. [+1.00]
.
I I
:
t
, - . -- ..
. . .
'
,
SENIOR REACTOR OPERATOR Page116 REFERENCE: 1. LPI 013-07 PG 30 2. 215004K402 (3.4/3.5) 215004K402 ..(KA's) ANSWER: 041 (1.00) b.[+1.00] REFERENCE: 1. LP 013 07 TABLE 2, LO 6 '
, .2. 215004K401 (3.7/3.7)
215004K401 ...(KA's) ANSWER: 042 (1.00) b.:[+1.00) : REFERENCE: ;
>l. LP 044-05 TABLE 1 L '2. 295020K203 (3.1/3.3).
L- 295020K203 ..(KA's)- L ANSWER: 043 (1.00)
'
g a.[+1.00]'
: REFERENCE:~
r 1.- LP " ALTERNATE LEVEL CONTROL" PAGE 21, LO 4,6 i 2,- 295031K101 (4.6/4.7)
*
295031K101 ..(KA's) I I
. - _- . .- . . - [ - .
SEisl9R REACTOR OPERATOR Pagell7 ANSWER: 044 (1.00) b.[+1.00) REFERENCE: 1. LP FIG 3 051-07 2. 241000A107 (3.8/3.7)
,
241000A107 . . (KA's) ANSh!R: 045 (1.00) d.[+1.00)
'i REFERENCE:
1. LP 048-08 pg 42, LO 3 , 2, 241000K602 (3.5/3.7) 241000K602 . . (KA's) c ANSWER: 046 -(1.00)
' d. [+1.00) -
REFERENCE: ' g ' l. LP 036-05 pg 23 - 2. 295009G005 (3.7/3.6) i 295009G005 . . (KA's) t
'
ANSWER: 047- (1.00) b. [+1.00];
..m_--- - _ - _ _ _ -.__-m___ _ -_- a+.-- _ _ - -
_ _ _ _
, '
' SEN1OR REACTOR OPERATOR Pagell8
' REFERENCE:
1. SA-AP.ZZ-002 pg 17 2. 294001A103 (2.7/3.7) 294001A103 ..(KA's) ANSWu b48 (1.00) c. [+1.00] REFERENCE:
'I. LO HC.0P-AP.ZZ 0002 REV 8 PG 20 2. T/S 3/4'0-2 3. 294001A102(4.2/4.2)
'
?
294001A102 ..(KA's) t i ANSWER: 049 (1.00) f d. [+1.00] REFERENCE: 1. 'T/S, pg'l-7, 3/4 6-47 2. 295035K101-(3.9/4.2) 295035K101 ;..(KA's) ,
:
050 (1.00)
'
LANSWER:'
.. d. ' [+1.00)
REFERENCE: 1
,
j
, :1. E0P OP-EO.ZZ-101 2. 295037K104 (3.4/3.6)
295037K104' ..(KA's)
. +y .-, -- e+-- ,pr-.- g , +a - y e- w -
. . - - _ . - . . . . *
SENIOR REACTOR OPERATOR Pagel19
'
,
ANSWER: 051 (1.00)
,
a.'[+1.00) REFERENCE: 1. LP 044 05 Table 1 3. 296032K303 (3.8/3.9) 295032K303 ..(KA's)
' ANSWER: - 052 (1.00)
b. [+1.00) REFERENCE:
'I.; LP 044-05 Table 1 ~2. ~295032K303 (3.8/3.9)
295032K303' ..(KA's) ! l l ANSWER: 053 (1.00) l
'b. [+1.00]-
REFERENCE: 1. E0P 102A, Primary Containment Control ! 20 295030G011;(4.3/4.5) ,
!
'
295030G011 ..(KA's)
4 ' i' ANSWER: 054. (1.00)
' 'a.=[+1.00] -
l l {
1 l-i i
-
- - ,
'
' SEN1OR REACTOR F iRATOR Pagel20 REFERENCE: 1. T/S 3/4 2.3.
2. 295005G004 (2.7/3.7) 295005G004 ..(KA's) ANSWER: 055 (1.00) b.[+1.00] REFERENCE: 1. LP 047-03 pg 11 and Table 1 2. 295032K207 (3.6/3.8) 295032K207 ..(KA's) ANSWER: 056 (1.00)- c. [+1.00) REFERENCE: 1. SA-AP.ZZ 046 (Q) pg. 4 2. 29400lK103 (3.7/3.7) 29400lK103 ..(KA's) ANSWER: 057; (1.00) b.'[+1.00) REFERENCE: 1. OP-AP.ZZ-102 "Use of Operations Department-Procedures", pg. 3 2. -294001Alll (3.3/4.3).
'294001A111 .. KA's)
;
__ _ . _ . . SE,NIOR REACTOR OPERATOR Pagel21 ANSWER: 058 (1.00) d. [+1.00) REFERENCE: 1. LP 029 UTILITY 2. 295003A103 (4.4/4.4) 295003A103 ..(KA's) a f I ANSWER: 059 (1.00) c. [+1'.00] ,
' REFERENCE: '
il. T/S 1.7 Utility 2. 215004K501 (2.6/2.6)_ 215004K501 '..(KA's)
~
'
ANSWER: 060 (l'.00)
'
d.-(+1,00) REFERENCE:
:le- OP-AB.ZZ-105 UTILITY 2i 295022K301:(3.7/3.9)
295022K301 ..(KA's) ,
* . ANSWER:' 061 (1.00)
d.[+1.00]-
,
b _-_ ._-- __-_ - _ _ ____ _ _ , , , _ I
.. . _ , . . _ . . -. - . .
SENIOR REACTOR OPERATOR Pagel22 REFERENCE: ! I LP 026-09 PG 50 UTILITY
'
1.
2. 206000K419-(3.7/3.8) ) 206000K419 ..(KA's) J ANSWER: 062 (1.00) b.-[+1.00) REFERENCE: 1. LP 096A 01 pg 36 2. 295033A108 (3.6/3.8) , 295033A1081 ..(KA's)
.
ANSWER::- 063-(1.00)
-d. [+1.00)
>
REFERENCE:
'l ~. LP 050 UTILITY 2. 212000A210 (3.6/3.8) !
212000A210 - ..(KA's)
.
'
'
ANSWER: 064 (1.00)
:a'[+1.00) i ' - REFERENCE::
16- LP 033-05 PG 21 2; .22300lG007 (3.7/3.8)
'
223001G007- ..(KA's)
,
I
- . , _ . , , , -w- a
_. _ _ _ - . _. _ __ SENJhRRENCTOROPERATOR Pagel23
.
ANSWER:' 065 (1,00) :
;
d.(+1.00) < REFERENCE: 1. LP 034-07 PG 24 2. 295038K302 (3.9/4.2) 295038K30'c ..(KA's)
. ANSWER:- 066 (1.00)
b.-(+1.00) REFERENCE: 1. E0P 101 CAUTION,-E0P 201 CAUTION, l 2. 295009G007 (3.4/3.7) : 295009G007 ..(KA's) .
-
l; .. ANSWER: . 067 (1.00) l
' .d.[+1.00] -
l L REFERENCE:.' ! L-
'l . ~ LP 002 PG 31 UTILITY '
2. 295031K209 (3.3/3.4)
,295031K209 ..(KA's)
ANSWER: 068 -(1.00) c.[+1.00)
, .e ,_,s . e. , . , - , . . _ ,
- , - - . . . . . . . . - . . _ , , , , . . . . , , , , . . . -
SEN1'RO REACTOR' OPERATOR Pagel24 REFERENCE: 1. .LP 31-05 2. TECH SPECS B3/4 6-10 3. 295012G004 (2.9/3.8) 295012G004 ..(KA's) ANSWER: 069' (1.00)
"
b.[+1.00)- _ _
' REFERENCE:
' 1. LP-059 05 PG27 2. 295009A101 (3.9/3.9) 295000A101 ..(KA's)
. ANSWER: 070 (1.00)
_ a..[+1.00) _ REFERENCE: _ _ 1. LP-059-05 PG-25 2. 29500lK607 (3.8/3.8) 295001K607 ..(KA's)
-ANSWER: 071 -(1.00) -
c. [+1.00) _ REFERENCE: 1. LP-030-07 PG B2 UTILITY 2.- 295005A207 (3.5/3.6) 295005A207 ..(KA's)
. -
SENIOR REACTOR OPERATOR Pagel25 ANSWER: 072 (1.00) d. [+1.00) REFERENCE: 1. OP-E0.ZZ-101, RC/Q-16 2. 295015G012 (3.7/4.4) 295015G012 ..(KA's) ANSWER:- 073 (1.00) c.[+1.00) REFERENCE:. 1. .LP 113-01, LO h.1 2. 294001K102 (3.9/4.5)
.29400lK102 ..(KA's) . ANSWER:- 074- (1.00)
b. [+1.00) REFERENCE: l '. 10 CFR 26 2. 294001A103 (2.7/3.7)- 294001A103 ..(KA's) ANSWER:'- - 075 (1.00) d. [+1.00]
- ,-
.
i SENIOR REACTOR OPERATOR Pagel26
' REFERENCE: '. l . SA-AP.ZZ-002(Q) REV 13 PG 22 2.- 294001Kil6 (3.5/3.8)
29400lKil6 ..(KA's) ANSWER: 076 (1.00) c.[.00] REFERENCE: 1. OP-AP.ZZ-109(Q) ATTACHMENT 1 PG 3 2. 29400lK107 (3.3/3.6) 29400lKiti ..(KA's) ANSWER: 077 (1.00) b..[+1.00] REFERENCE: 1. OP-AB.ZZ-300 PG 3 2, 29500lG010 (3.8/3.7) 295001G010 -..(KA's)
, : ANSWER: . :078 (1.00)
c.-[+1.00] REFERENCE: p., 1. OP-AB.ZZ-300 PG 2 2, 29500lK104 (2.5/3.3).
295001K104 ..(KA's)
"
___
. - . .-.- - -. .- - - _. - .
SENIOR REACTOR OPERATOR Pagel27 i I-l ANSWER: 079 (1.00) d. [+1'.00]
- REFERENCE: ,
1. 10 CFR 20 2. 29400lK103 (3.3/3.8)
,, ~29400lK103 . . (KA's) ' ANSWER: 080 (1.00) ,
c.'[+1.00) REFERENCE:-
*
1. OP-AB.ZZ-0131 PG 1 l 2. 295019A202 (3.6/3.7) 295019A202 . . (KA's)
> - ANSWER: 081 (1.00) . .b. [+1.00] .
1 REFERENCE: 1. .LP-103-02 FIG 1 2.'29400lKil5(3.4/3.8) .; 29400lKil5 . .
(KA's)
ANSWER: 082 (1.00) L b. [+1.00] h
.
>
w r--' *
--e - - r
_ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ - _ _
. .. . 'SEN!bRREACTOROPERATOR Pagel28 REFERENCE:
- - 1. LP 113-01 LO 2 2. . 294001A106 (3.4/3.6) 294001A106 ..(KA's)
' ANSWER: 083 (1.00) _ c.[+1.00]
- - . REFERENCE:
- 1. LP 113-01 pg 12, LO 12-2.'294001A103 (2.7/3.7) _ 294001A103 ..(KA's) _ - , . ANSWER: 084 (1.00) , a.(+1.00] REFERENCE: 1; . TECH SPEC PG l-10 TABLE 1.1 -
'2. 294001A108.(3.1/3.6)
294001A108 ..(KA's)
-
ANSWER: 085 (1.00)' _
'c..[+1.00] :- ' REFERENCE:
T_ 1. TECH SPEC 3/4 4.9.3 PG 3/4 4-15 l_ 2. 294001All4 (2.9/3.4) 294001All4- ..(KA's) -
:
_
$
_ _ _ _ _ _ _. - . - - _ _ _ . _ . - . _ _ - . - - . _ . . _ - . . _ - - - _ _ _ _ . - - - _ . -_._.-_._-_n --- - - - - ~ . . _ - - . - _ _ - _ _ _ _ _
'
.
SENIORREkCTOROPERATOR Pagel29
- -
ANSWER: 086 (1.00) d.[+1.00) REFERENCE: 1. LP-113F-01 PG 10 2. 294001A109 (3.3/4.2)
'294001A109 ..(KA's) ' ANSWER: 087. (1.00)
c.-{+1.00] REFERENCE:
,
1. SA-AP.ZZ-015 UTILITY 2. 29400lK102 (3.9/4.5) 29400lK102' ..(KA's)
-ANSWER: 088 (1.00)
b.-[+1.00)
' REFERENCE:
1. OP;AB.ZZ-121(Q) 2. 295026G010 (4.0/3.8) 295026G010 ..(KA's) ANSWERi 089 (1.00) d.[+1.00)
, t
- SENIOR :REdCTOR OPERATOR ' Pagel30 REFERENCE:- ' ' l'.---OP-AB.ZZ-135 PG 2 ~ 2.( 295003A202 :(4.2/4.3) -
2h5'000A202 ..(KA's)-
,
ANSWERi: 090 -: ' (1.00) :
' b.? [+1.00] ,
g REFERENCE:
'
h 1.'[0P-ABZZ-201~PGla < L2. 1295020G010 '. (3.6/3. 5)'
>> ~295020G010 T ..(KA's)-. -
, ,
ANSWER:- 091 .
(l'.00)/ :n .: a.1[+1.00] 2 REFEREN , > :1. ~0P-E0.ZZ-102A PGL19'
T2i ; 295030K207!(3.5/3;8) L295030K207; '-..(KA's) i
.
i e , , YJANSWERi 092 (1.00) .
'
d. [+1.00]__
,
REFERENCE:'
'
1. - LP-~101-03'PG 16.
_. 2. '295017K201.(2.8/3.3) I L295017K201 - ..(KA's)
. .
.. ~
3. . - .
, ;- SEN15RRENCTOROPERATOR' Pagel31 ,
n . L
,
l' -h
-
-
.
'i -
: ANSWER: -093 (1 00)
.
[
.
c i a . L [+1. 00'- a-REFERENCE:
-lb=OP-E02ZZ-101 RPV CONTROL- ;2.
295013G012.(3.6/4.2) ; y + r
-
~ 295013G012 '
,..(KA's).
4. .
- '
y : l Oi :JANSWER':;;094?.l(1.00) .
-u
~ "., u ,u - 'cj[+1.00)
.
, l
-
c ,_a s .
,
f l l - i
.' * 1:RE'FERENCE:)
m M "11..;LP-002-05 PG16.. ..
'
lcl,3 '2.- 295008K304g(3 3/3.5) u w.
-
, .
Y'Nd295008K304 . (KA's) , Wyy
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dm .-
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pj}R-d. . . . ; 691.'00] , k+ l9 t ;
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.t' 0 REFERENCE: '
!
#- 1
.
'g s O' l4 ' l.(-LP-109 UTILITY e f 3, ;2. .
295014K105.(3.7/4.2) 'I . 2% g W gj95014K105 ..(KA's).
,,q ,
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n. : , .'{w: ' ' , yffm[ANSWERP 7 096 (1.00)
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4; . ..; , ll WSENIOR REACTOR;0PERATOR Pagel32 !
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g 1
'
REFERENCE:, I
. . .
-
1.w OP-AB.ZZ-300 REV 1'PG 2 -
'
s E ,: 2. . 295001K104:(2.5/3.3)
~ 29500lK104? ..(KA's)
'
-
ANSWER: 097c (1.00)
=
;
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'
l REFERENCE: y: 3,, .! N,p .'1.
'
-
iP 024 UTILITY LO 7 lQ';2.;295037A104(4.5/4.5), 1: n .
, .J e 295037A104 ..(KA's)' l e
1.'
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I 3 i!:;ph . F ,ANSWCR* 0981 (1.00)-
- .
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['gj i % ~a,l[+1.00)- o t i; ~Jf.: ' .i
- . REFERENCE:?
7.Il 'l. =0P-E0.ZZ-101 ED-3 UTILITY.
2.1 : 295037G012'(3.9/4.6)
; '
. x
- j L295037G012' .
- . (KA', s )
'
Ew , u
:; . ANSWER: l '099 (1.00) - b.- [tl .00] '-
-t . ' REFERENCE: 1o
, ' l.: LP-30'2-129-05 PG 21 UTILITY , w 12( 1295031K101L(4.6/4.7) ;
\ ' , ,
295031K101 ..(KA's). .i
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_ _ _ _ _ . _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . - . _ _ _ _ _ _ . _ _ - _ _ _ . _ . . _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ - - _ _ _ .
- - . . - - . . . - . .. ..
SENIOR REACTOR OPERATOR. Pagel33
, <
\ jL VfANSWER: =100 (1.00)-
>
. . .
-* kc.?[+C00).
+, tE y. ,
'
- REFERENCE: -
./ :1;. OP-EP.ZZ.101, "RPV Control" ,' 2. 295037K307-(4.2/4.3)
' N' 295037K307,- ..(KA's)
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} (********** END Of EXAMINATION **********)
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Attachment 3 5 o % Facility Comments on Written Examination o
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> .Pubhc Service Electric and Gas Cornpany -P.OI Box 236 Hancocxs Bridge, New Jersey 08038 ,
Hope Creek Generating Station
,
September 26, 1990
# "' NTC-90-3197 :
Mr. T.. Timothy Martin
,i Regional' Administrator.
LU.S. Nuclear ~ Regulatory Commission Region I-
> 475 Allendale Road- ' ,
I King of Prussia', .'. PA . -- 19 4 06 p' S
Dear Mr. Martin:
-
' EXAMINATION REV.!EW COMMENTS - HOPE CREEK LICENSE EXAMINATION zy , " .
~
,
> Attachod:are: comments on two questions used on the written.
examination administered;at Hope' Creek-Generating Station on
, = September 24,.1990. . These' comments have been developed.following: , '
i1 theEexamination.. They 7 are: in; addition - to comments . provided by'
' :q LMessrs.(Jones, O.! Mal ~ ley, Bauer, Martin,.and Gott during the W pre-examinationnreviewJconducted on September.19, 1990.
E, :The1following'. format has been used to document specific comments: 5 o
,J NRClquestion,canswer and referencor
~
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, ;A. *
'
m ' S / B. facility comment including a recorum.ndation for. resolution;; . y
,
f, 4 Lan'd-
, , 1C. support documentation.
Jh . ... Theseccomments are presentod-in the
. '
order as originn)1y- > h '. ,. numbered on theLSRO examination,.but L.sy are applicable to both.
nl l
'the ROLand SRO cxaminations. The RO examination queution number t,
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. .- . * . .-
Mr. T. Timothy Kartin
. -2- 9/26/90 If you-have any questions, comments, or need additional information, please call G. Mecchi, (609) 339-3857, or W. Gott, (609) 339-3769. They will provide the requested information or will see that you are contacted by the appropriate person.
Sincerely yours, lr C. P. Johnson 3 he ($
(Acting) General Manager -
Hope Creek Operations l WTG:al Attachments
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A. NRC Ouesti6n, Answer and Reference SENIOR REACTOR OPERATOR Page 44
!
QUESTION:. 041 (1.00)
WHICH ONE (1) of the following SRM rod blocks / scrams are bypassed with its associated IRMs on Range 3? ; a. SRM. Inoperable
=b. SRM detector wrong position
, i c. SRM^ upscale reading 1.5 EE +5
?, d. SRM detector not full in , ANSWER: 041 -(1.00) b. (+1.00] L REFERENCE: l'. .
'LP 013-07 TABLE 2, LO 6 215004K401 (3.7/3.7) '
L 2.
L 215004K401 ..(KA's) B.- Facility Comments l NOTE: BOTH RO EXAM OUESTION #40'AND SRO EXAM OUESTION #41 ARE THE SAME OUESTION. ! During the pre-examination review,.it was agreed that the wording of. item b, the correct answer on the, answer key, be changed from "SRM detector. wrong position" to "SRM. detector
.
not. full in". While "SRM detector wrong position" is used in the lesson plan (LP-013-07), "SRM detector'not full 11n" - is the more common usage at Hope Creek'and is'the wording fo- 3 in Tech. Specs.
In any. case, "SRM detector not full in" appears'to have been-inadvertently. included as the distractor at item d, and resulted in two correct answers for this question..
.
It is recommended that both answers b and d be accepted as Correct.
C. Suncort Documentation HOPE CREEK TECHNICAL SPECIFICATIONS 3.3.6 TABLE'3.3.6-1 0 3a on page 3/4 3-57 HOPE CREEK TECHNICAL SPECIFICATIONS 3.3.6 TABLE 3.3.6-2 0 4a on page 3/4 3-59 HOPE CREEK TECHNICAL SPECIFICATIONS 3.3.6 TABLE 4.3.6-1 0 3a on page 3/4 3-60 Attachment 1 s Page 1 of 4
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R - TRIP FINICTION L OPERATIONr.I.
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' 6. ' REACTOR COOLANT SYSTEN SECISCNEATION FLOW'.
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-TABLE 4.3.6-l'
CONTROL R00 BLOCK INSTRUE NTATION SURVEILLANCE REQUIREMEWIS
' - n CHAfMIEL OPERAIIONAL--
! E CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH 3 -TRIP FUNCTION. * -
' CHECK ~7EST CALIBRATION I 'I SURVEILLANCE REQUIRED l 1. ROD BLOCK MONITOR /
i a. Upscale- -NA Z(c)(d)'-(c) 34 3
'Z(c)(d)',(c)
' b.- Inoperative- .NA NA 1* c. Downscale NA Z(c)(d) (c) SA 1*
~
2. APRM-l a. ' Flow Biased Neutron Flux - - i . Upscale NA S/U(b).f ,Q SA 1
' b'. Inoperative NA. S/O ,Q- NA' '1, 2, 5 c. Downscale- NA: 5/UIbI,Q SA: 1 d. -Neutron Flux - Upscale, Startup- -NA ~ ,
S/U ,Q 5A 2, 5 4gi 3. SOURCE RANGE N)MITORS.
'* $ T a'. NA_ S/U(b) W , NA 2, 5
^i 8 b. Upscale NA S/U ,W SA 2, 5
,Rg c. Inoperative NA- W NA 2, 5 , d. Downscale 5/U(b),W S/U
- ,n NA
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4. INTERMEDIATE' RANGE MONITORS a. Detector.not full in MA S/U(b) NA 2, 5 b. Upscale NA 5/U(b),W- ,W SA 2, 5 c. Inoperative NA S/U ,W NA 2, 5
. d. Downscale .NA S/U ,W SA 2, 5 ,
5. SCRA!4 DISCHARGE VOLtME-a. ' Water. Level-High (Float Switch)-
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i i A.- NRC Ouestio'n. Answer and Reference Page 47- 1 SENIOR REACTOR OPERATOR ) QUESTION: 044 (1.00) l
.
l I GIVEN that:
- Load selector = 1067 MWe
- Ioad limit = 107%
- Max.Jcomb. limit = 113%
- Throttle pres. = 947 psig
A" pressure reg. = 920 psig I /
- The "A" EHC pressure regulator setpoint fails downscale. (See
'
. Attachment #4 point #1.') WHICH ONE (1) of-the following is the n resulting bypass valve steam flow signal of the EHC logic system?
,
.
a. 0% steam. flow a
.b. -6%isteam flow _j . c .- 7% steam flow
'- 'd.. .13% steam flow ANSWER: 044 (1. 0 0) -
, 'b. (+1.00)- l ' : REFERENCE:
1. LP' FIG 3 051-07 ' 2. 241000A107: (3.8/3.7)
'
is B. Facility Comments'
>
NOTE: OUESTION 844 IS THE SAME ON BOTH THE RO AND'SRO e EXAMS.
-Duringsthe pre-examination review,.this quastion.was-discussed in: terms of what-was meant by a " regulator u J~ setpoint" failing downscale-and where.it was. sensed in Attachment 4. This prompted 1the additioncof:the arrow at ,
'
s ' point #1. 'The reviewers did.not-identify.that;the Load .;
.
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LSelector was -set to less thanL100% Eof turbine ' generator
[" rated 01oad,;1118:MWe. This~ oversight may have occurred- l
' because the reviewers; failed both to identify this controll as" Load Set, as does the figure.on Attachment 4', and to'- account for its.. significance'at the stated.value,:1.e., Load Selector = 1067.MWe. .Normally, Load Set is adjusted'to 1150 2 MWe:after the generator is put on line'-.
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.
.Under stated conditions, analysis reveals that on the given failuto-the control Valve demand signal would be limited by
,
Attachment 2-
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Page 1-of 4
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W i' B. Facility' Comments -(Continued): the; Load Selector.(Load Set) to a maximum value of 100%;or I o- less. 'If limited-to 100% and with the Maximum Combined Flow tc Limiter at 113%,.the Turbine Bypass. Valves'would receive a 13% f - steam flow signal. If' limited to less than 100% steam flow it signalLto the Turbine Control valves, the Bypass Valves would ' '
- receive a' steam flow signal greater than 13%. The largest value l
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given;as,an answer is 13%. t e; L lt is recommended that answer d, 13%,'be accepted as the !~ answer that is closest to a reasonable value'that can be E .
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calculated:for the given conditions.
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-- C ; Suonort Documentation i ls t It -HC.OP-IO.ZZ-0003(Q)LSECTION 5.4.9 ,
ATTACHMENTi#4,OF THE EXAM ..
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CAUTION f_4'.9
' Failure to establish load set at the proper :value with a subsequent -loss -
of stator. water cooling.will violate
; the runback' analysis and may result in substantial generator. damage.
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[. 5.4.9 Increase the-IDAD'SETJindicator' !
to as close as possible to, hut-
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t not greater than,.1150. MEGAWATTS. i
. 8
[ < .5.4.10 CLOSE(tho'following:- ,! L f l[ jp '7 A. HV-1026 STM LEAD .S/UJ(1-Pb) - - W, LB. <HV-1013'A,B,C,D MN STM VLV TR SEAT
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C. HV-1015. CONT VLV BFR SEATa(1 Pb)'.
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01 D. HV-.1017A/B. STEAM LEAD 1&2 (l'.Pb).
p , li< -E. "HV-F018B' STEAM' LEAD 3' y ,
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ws . - JW .n .5.4.11 PLACE / ENSURE #4' STEAM LEAD DRAIN ? ><
' '
HV-F018A is in. AUTO; '
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. . . ..
1 CAUTION'5L4.13.
. Power-changes-shall be?done with'the1 [ ,f guidance'of the:SenioriReactor.
O H Supervisor. , j M,/ --
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4.4 s 5.4'.12 Continue control rod withdrawal;IAW [
'
%g'ig j the Rod Pull Listing. i ~ . .e , f/NT 5.4.13 At.'approximately,15% load, perform the % x I4s, following: Transfer Reactor Feed
.
!* A.
! Pump , A(B,C) . P101' Control to the Master Level controller IAW OP-SO.AE-001(Q), Feedwater
:
System Operation.
< Hope Creek Page 28.of 42 Rev. 24 Attachment 2
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NRC' Resolution of Facility Conments ,
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' Project Number M54835
: 15 1 Pacific Northwest iaboratories internal Distribution ,
, ll :i- file /LB- t one October 9, 1990-To CHIEF EXAMINER, HOPE CREEK, SADA PULLANI
>
' rvom . .J.J.-MUTH-
.,..
n subject EXAM COMMENTS' RESOLUTION !
, t '1. rt0 question'. #40, and SRO. question #41 are the same. _ Utility comment request. ' -
that answer B and'O be accepted'due to their similarity.
'- RESOLUTION: Comment' accepted ^
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2. R0. question #44 and Sr.0 questior #44 are the same. Utility comment request ; that:0 be accepted.bec m.e-it is tie closest to the correct answer. '
. .
RESOLUTION: Comment not accepted and-qu% tion is deleted because there is N0 [ 0* correct-answer..
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p "" l l l i Attachment 5 '
!
.-Simulation Facility Report
, ,
Facility Licensee: Public Service. Gas and Electric Company I l Facil.ity Docket No. :'50-354
~
Operating Test'. Administered on: September 26 - October 3, 1990
, , ,
This. formisfused only to report' observations. These observations do not constitute audit or inspection findings and are not, without further verifi - cation and: review, indicative-of non-compliance with 10 CFR 55.45(b). These! , tobservat'i'o ns do notief fect: NRC certification or approval of the simulation
~ '
facility othen than:to' provide i'nformation which may be used in~ future evalu- 2 ations. No licensee action isf required. in: response to these observations,
'
During the donduct J of the simulatoriportion of the operating tests, the
,
following item was observed:
.
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:
Item' ' Description j
Ourin'g Examination Scenario 5-1; conducted on September 28,
' : Va; tor L Pro'ce ssor?
1990, thelsimulator died for1a sfew minutes ~ The cause was "
- .'
determined to be a ffreeze-up of the. Vector: Processor in: the. J Sp , simulator. .The simulator'could be reset;immediately and 5 thelexamination continuedcwithout any: problem.
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