ML20059L163
| ML20059L163 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 09/07/1990 |
| From: | Jordan M, Rau E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20059L160 | List: |
| References | |
| 50-440-OL-90-02, 50-440-OL-90-2, NUDOCS 9009260193 | |
| Download: ML20059L163 (221) | |
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U.S. NUCLEAR REGULATORY COMMISSION REGION III l Report No. 50-440/0L-90-02 I Docket Na. 50-440 l Licensee: Centerior Energy - The Cleveland Electric Illuminating Company Facility Name: Perry Nuclear Power Plant
- Examination Administered At: Perry Nuclear Power Plant .1 Examination Conducted: August 6-10, 1990 I l
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- l. Chief Examiner: h 4 L E. D. (Rau ;
01')0 Approved By: _ / // V . < i
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M.J.ffordan, Chief I l Operating Licensing :ection 1 Examination Summary: Examination administered on Auoust 6-10. 1990. (Recort No. 50-440/0L-90-02) . Written and operating examinations were administered to seven reactor operator 4 and four senior reactor operator candidates. Results: All candidates passed the examinations.
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1 o I ' i 9009260193 900907 PDR ADOCK 05000440-V PNU _. l
DETAILS 1.0 Introduction and Overview The NRC examiners administered replacement examinations to seven (7) reactor operator (RO)' applicants; one (1) SRO instant applicant; and-three (3)-SRO upgrade applicants. The examinations were administered in accordance with NVREG 1021. Examiner Standards, Rev. 5. Prior to administration of the examinations, the written examination was reviewed with personnel from the training department. The. scenarios used for the operating portion of the examinations were prepared by the NRC and run on the plant specific simulator prior to administration. All facility individuals involved with the review of the examination materials signed security agreements to ensure that there was no compromise of the examination. 2.0 persons Contacted
*D. Bauguess, Licensed Instructor
*D. Igyarto, Manager of Training
*M. Wesley,- OTU Supervisor
*H. Hegrat, Licensing and Compliance
*J. McHugh, License Training Coordinator
*K. Matheny, Simulator Project Manager
*C Frank, Licensing Engineer M. Haskins, Non-Licensed Instructor N. Johnson, Licensed Instructor
- Individuals that were present at the exit on August 10, 1990.
3.0 Examination Related Findinas and Conclusions 3.1 Summary of Results R0 SRO Pass / Fail Pass / Fail Written 7/0 4/0 Operating 7/0 4/0 Overall 7/0 4/0 3.2 Operatina Examination The following is a summary of generic strengths and weaknesses noted on the operating tests. This information is being provided to aid the licensee in upgrading license and requalificat ~ n training programs. No licensee response is required. 2 1
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Ih Strenaths t The candidates demonstrated good knowledge or strengths in the following areas:
- a. Knowledge of the in plant location and operation of equipment
- b. Knowledge and use of procedures
- c. Shift turn-over
- d. Communications EtAk.nesses The candidates demonstrated lack of knowledge or deficiencies in the f following areas: ;
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- a. Candidates had difficulty operating main steam isolation valves for alternate depressurization -
- b. Candidates did not make aggressive attempts to regain equipment :'
that had failed
- c. Knowledge of operational event effects on thermal limits +
- d. Peerational reactor ' physic.s ,
- e. ,3 3.4.1.1 LCO interpretation +
- f. 10 CFP.f>3 license conditions .
5 3.3 Written Examinatiqns i The following is a suinmary of the generic strengths and weaknesses noted ; from the grading of the written examinations. This information is being j provided to aid the licensee in upgrading license and requalification
- training programs. No licensee response is required.
l Strenaths The candidates demonstrated good knowledge in the following areas: l a. Knowledge of procedure immediate actions Weaknesses The candidates demonstrated lack of knowledge in the following areas:
- a. MSIV leakage control system operations
- b. Protective action recommendations 1 1 - c. Confined space entry requirements e d. RPS/ turbine valve interrelationships ;
L e. Chemistry tech specs
- f. Battery tech specs >
- g. Clinton LER 3.4 Facility Comments on Written Examination l-i 3
Approximately 11 post-n am comments were received for both the RO and l SRO written examinatins. All comments were reviewed and dispositioned i as shown in Enclosure 4. Breakdown of comment resolution is as follows- i 3 questions - answer key changed; 7 questions - two answers accepted and k 1 question deleted because of no correct answer. The reasons for each i examination comment are varied and do not imply that the facility pre-examination review was not adequate. ) 3.5 Plant Soccific Simulation Fa .f 3 A number of problems with simulator fidelity were identified as a result of the examination process. These problems are discussed in Attachment s I to this report, j 4.0 Other Items of Concern Several other items of concert were noted by the examiners and were j presented at the exit.
- a. Use of combustible material as enclosures for seismic detectors in auxiliary building.
- b. Designation of entire containment as a dress-out area. Although ;
t several recent operational events have contributed to this ! designation, maintaining the entire containment as a dress-out ; area hampers operations and contributes to rad waste,
- c. Consider implementing a design change to modify RCIC Initiation-Turbine Trip Circuit to function in the manner of other BWR 6's i (IE:15 see TO). i l- 5.0 Exit Meetina An exit meeting was conducted on-August 10, 1990 following the administration of the examinations. The licensee re)resentatives that attended the meeting are listed in Section 2.0 of t11s report, i There were no problems with access to the plant and Operations personnel were cooperative. The generic strengths and weaknesses noted on the examination were presented (see Sections 3.2 and 3.3 of this re) ort).
The problems with simulator fidelity (See Attachment 1), and otler items of concern (See Section 4.0) were also discussed. 1
Attachment:
- 1. Simulation facilit Report 4
4 SIMULATION FACILITY REPORT Facility Licensee: Perry Nuclear Power Plant Facility Licensee Docket No. 50-440 Operating Tests Administered At: Perry Nuclear Power Plant During the conduct of the simulator oortion of the operating tests, the following items were observed: 11[M DESCRIPTION
- 1. MSIV's MSIV's exhibited excessive seat leakage
- 2. Rod Drift Alarm Spurious rod drift alarms during rod movement 3.- DC Bus Failure DC Bus failure not modzled correctly Malfunctions (IE: Some valves VPI indicates energized duringinstallationoffailures)
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.. 2 OlfESTION: 055 (SRO) 060 (RO)
POINTS: 1.00 SELECT the operator action required to be taken 1e procedure if the "A" recirculation pump trips while. operating at rated power. Pland conditions following the recirculation pump trip are listed below. Figure A006, "Pover to Flov Operating Hap," and ONI-B33-2, Attachment 1, " Thermal Power versus Core !
; Flow" ate provided as Attachment 8.
PLANT CONDITIONS APTER TE "A" RECIRCULATION PUMP TRIPS Reactor Pover 45% vith 6% bandvidth oscillations. Core flow: 42 M1bm/bo j
- a. Arm and depress the RPS manual scram pushbuttons, j
- b. No immediate operator actions are required.
- c. Insert control rods in reverse sequence to lover reactor power below the limits of ONI-B33-2, Attachment 1.
- d. Insert the CRAM rods to lower reactor power below the limits of ONI-B33-2, Attachment 1.
ANSVER: d. ! i COMMENT: Both ansvers "b." and "d." are correct depending on-the l reference graph that is used. On the Power to Flov map, with the oscillations, you are in the Shaded Region, thus D is the lE correct answer. The ONI curve indicates that you are-not in the shaded region; thus "b." is correct. This discrepancy is due to an improper ONI curve which has already been corrected in the plant effective (-4-90. RECOMMENDATION: Accept ansvers "b." and "d." REFERRNCE: Graphs attached NRC RESOLUTION: Concur. Accept answers "b" and "d". I
T QUESTION: 052 (SRO) 056 (RO) POINTS: 1.00 Given the following initial Offgas system alignment: INITIAL OFFGAS LINEUP Charcoal adsorber control switch: " BYPASS". Charcoal adsorber bypass valve (IN64-F045): Open. Adsorber inlet and outlet valves (IN64-F051A-D and IN64-F053A/B): Closed.
"A" Steam Jet Air Ejector (SJAE): In service.
SELECT the expected realignment of the offgas system valves when the following radiation monitor indications occur. RADIATION HONITOR INDICATIONS The K601A post-treat radiation moni tor is- alarming and i ts "HI" indicator light is illuminated. The.K601B post-treat radiation monitor is alarming and its "DOVNSCALE/INOP" indicator light le illuminated.
- a. Charcoal adsorber bypass valve (IN64-F045): .Open.
.Adsorber inlet and outlet valves: Closed.
(IN64-F051A-D and 1N64-F053A/B) Offgas discharge valve (1N64-F632): Open.
- b. Charcoal adsorber b/ pass valve (IN64-F045): Closed.
Adsorber inlet and outlet valves Open. (IN64-F051A-D and 1N64-F053A/B) Offgas discharge valve (1N64-F632): Open,
- c. Charcoal adsorber bypass valve (IN64-F045): Closed.
Adsorber inlet and outlet valves: Open. (IN64-F051A-D and IN64-F053A/B) Offgas discharge valve (IN64-F632): Closed.
- d. Charcoal adsorber bypass valve (1N64-F045): Open.
Adsorber inlet and outlet valves: Closed. (IN64-F051A-D and 1N64-F053A/B) Offgas discharge valve (IN64-F632): Closed. ANSVER: b. COMMENT: Vith no operation action, selection "a." is correct due to modo svitch in bypass (208-163 sheet 13). The operator vould be expected to realign system (ARI-P604-1 and S01-N64/62). RECOMMENDATION: Accept ansvers "a." and "b." since question does not specify with or without operator action.
REFERENCE:
208-163 sheet 13 1RI-P604-1 SOI-N64/62 NRC. RESOLUTION: Do not concur since operator action is not specified. Only answer "a" is acceptable. -Answer Key changed to Answer "a"'. ,,
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" LOUESTION: '061 (SRO)
, , , 065'(RO)
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POINTS: 1.00' t
. SELECT;theLoperator_ action that vill reestablish Control' Rod Drive Hydraulic lA (CRDH)~ drive differential pressare following a high dryvell pressure scram that ,
cannot be reset. CRDH pump "A" isl running. [ W' . ..
- a. -
Bypass the CRD' pump suction filters.
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Ts b. : Place the flov. control valve in " MANUAL" and open it. < ' l c, Fully close the-drive water pressure control valve. [; . [ 21 . Take tiie high dryvell pressure scram bypass switches to "MPASS" ~and reset. ' the scram and ' alternate rod insertion (ARI). y
. ANSWER: b.
l COM'MENTs.
' PEI B-13'provides geldance to start a 2nd CRD pump.<;This is c ' the only reference thc;' is to this' situation. Action "c."-
can be used to obtain the name results. RECOMHF.NDATION: Eliminate question
REFERENCE:
PEI B13
. .Mic RESOLUTION Concur ~ .
Question climinated. s c1 J 4
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- OUESTION: 063'(SRO)
F-- ' POINTS:. 1.00 L: ., h : L j SELECT the statement that describes a. condition that allovs energizing stub bus 3 XH12 during a'large break Loss of Coolant Accident (LOCA)_ condition. The _ ' C caution note from PEI-D23-3,.Section 5.0, "Special Operations" is reproduced + li: .below. t L CAUTION ; e -l F Before enerizing the' stub bus in the following steps, verify-RPV "
' level can be determined, and no ECCS pumps associated with the !
respective stub bus are providing adequate core cooling. I subsequent.to the stub bus energization event, when any of the L , ECCS pumps associated with the bypassed stub bus are providing
, adequate care cooling, then open the ISOLATING BRKR-EH1116(1214) t~
and plae; ..e BUS XH11f.12) LOCA BYPASS keylock'svitch in NORM.
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- a. The Lov. Pressure Core Spray (LPCS) pump is injecting at 6500 gpm, RPV vater level is -12 inches and stable. :
- b. The Residual fleat Removal' (RHR) pump "C" injecting at 5000 gpm, RPV vater' [
J _-level-is 44; inches and stabic. ' "
- c. The Lov Pressure Core Spray.(LPCS) pump is injecting at 3500.gpm,iRPV vater level-is 45 inches and increasing. t
- d. The Residual lleat Removal (RilR) pump "B is injecting at 7000 gpm, RPV l1 l- vater level is -5 inches and increasing. 1 5
- i. ANSVER: ~
a. F , COMMENT: "c." is correct per'PEI,813, section 3.3.3. . If level is above. j 0" you have adequate core cooling. Accept ansvers "a." and "c. RECOMMENDATION: ' j
, - REFERENCE ' PEI B13, Pg. 56JStep I?
"a." is correct. i h LNRC RESOLUTION: Concur. Accept answers "a" -
and "c". k i P
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fl -. . { 1 9 4 100ESTION: 011 (RO) 013 (SRO) POINTS: 1.00 Civen the fellowing surveillance data SURVEILLANCE DATA Recirculation pumps are in fast speed. 1B33-R613: 45 H1bm/hr. -r Jet pump data (Hlbm/hr): 1833-R611A, JP5: 1.6 , 1B33-R611B, JP10: 1.7 1B33-R611C, JP15: 1.7 IB33-R6110, JP20: 1.6 b APRM Flov readings (%): Div. la A: 39 E:- 39 Div. 2: Bi- 36. F: 36 ' Div. 3: C: 40 G: 41 Div. 4 - D: 38 H 38 SELECT the statement that describes the results or the performance of SVI-C51-T0026, "APRM Plow Biased Power-Flov Verification." Copies of A003,
" Percent Drive Flov vs. Total Core Flow" and SVI-C51-T0026 are-provided as '
Attachment 1.
- a. .Both div. 1 and both div. 3 APRHs are inoperable, t
- b. The "G" APRM is inoperable.
- c. Both div. 3 APRMs are inoperable. .
- d. Surveillance is satisfactory.
ANSVER: d. COM'iENT: Answer is incorrect RECOMMENDATION: Change ansver to selection "b." REFERENCES- SVI-C51-T0026 NRC RESOLUTION: Concur. Answer Key changed to "b". i
I . - g 1 b OUESTION: 019 (RO) ! 021 (SRO) , POINTS: 1.00 ' SELECT the statement that describes the effect on the Recirculation system for an actuation of the end of core life (EOC) interlock following a main turbi e trip.
- a. At less than 40% rated power, the lov frequency motor generator (LFMG) and the CD3 and CB4 breakers vill trip for each pump.
- b. At greater than 40% rated power, the CBS breaker vill trip, the lov frequency motor generator (LPMG) vill start, and the CB2 breaker vill close for each pump.
- c. At greater than 40% rated pover, the CB3 and CD4 breakers vill trip, the lov frequency motor generator (LFHG) vill start, and the CB2 breaker vill close for each pump.
- d. At less than 40% rated pover, the CBS breaker vill trip, the lov frequency motor generator (LFMG) vill start, and the CB2 breaker vill close for each pump.
ANSVER: b. COMMENT: The 3 and 4 breakers do receive a trip signal from E00 RPT as. documented on the attached prints. The 5 breaker vill trip as a results of an EOC RPT. RECOMMENDATION: Accept ansvers "b". and "c."
REFERENCE:
208-040 sh. All 208-015, sh. 27, 29 NRC RESOLUTION: Concur. Accept answers "b" and "c". m
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>OURSTION:~ 010 (RO) '
POINTS:- 1.00 i P i r ,
.f M i/ : SELECT the statement that describes a condition that must-be' met for the i @ /, Supervising Operator!at the controls to be-temporarily relieved for personal- -l p reasons. .;
a , y , a.- The Shift Supervisor's approval must be.obtained. '[
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"b. A. verbal.turnovot of status of vork in progress must.be completed. ;i LA i
g y- ic. The "30 Relief / Turnover Checklist" must be signed by the oncoming operator. l 4 ; Il '
- d. The oncoming operator signs'into the Unit Log as-Supervising Operator at ,[
the controls. i
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ANSVERA b.
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COMMENT: .One._ trainee was informed by the proctor that this relievingL i r person was not on shift.and that he came in from off-the. i
' street. In this case "c." is also correct as per 0AP 0103. l.
RECOMMENDATION . ' Accept ansvers "b." and "c." . 3. i*
REFERENCE:
DAP 0103'pg. 7'and 8
,, NRC RESOLUTION: ' Accept anspers "b" or '?c" for candidate GODA. . f o
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OllESTION: 004 (80 and SRO) POINTS: 1.00 SELECT the statement that describes the quarterly whsle body (VB) exposure limit for an individual with the exposure history-identified below as per PNPP administrative guidelines. EXPOSURE IIISTORY Ager : 25 years Lifetime exposure unknown NRC Form 4: not available, exposure history is undocumented
- a. 300 mrem.
- b. 1000 mrem.
- c. 1250 mrem.
- d. 3000 mrem.
ANSVER: a. COMMENT: Question and ansver are correct as written. Proctor vrote on board during exam that existing quarterly exposure was known to be 0 mrem. This information makes the question very confusing. If quarterly exposure is not documented, "a." is correct. If it is documented, "b." is correct. RECOMMENDATION: Accept ansvers "a." and "b."
REFERENCE:
PAP 0514, Pgs 4 and 5 NRC RESOLUTION: Concur. Accept answers "a" and "b".
. . . . . . . . _ ~.'....
L QUESTIONS. 086 (SRO) 088 (RO) _ POINTS: 1.00 SELECT the statement that describes the reason for requiring suppression pool-level to be greater than 5.25 feet prior to commencing emergency depressurization. 1: a. The Safety Relief Valves (SRVs) tailpipe vould be uncovered and discharge f m directly to containment, the containment design pressure could be exceeded. ?;! b. The submergence of the Safety Relief Valves (SRVs) is insufficient for f complete exhaust steam condensation and the containment design pressure $ could be exceeded. ?! M 1
- c. The reduced mass of water that is cleared from the tailpipe when the Safety Relief Valves (SRVs) is opened vould place excessive velocity loading on the tailpipe.
_ d. To provide sufficient back pressure on the Safety Relief Valves (SRVs) to ensure the SRVs vill reclose when the control switches are placed to "CLOSR". ANSVER: a.- COMMENT: The selections for "a." and "b." are very similar. The quencher is the part of the SRV that is uncovered, not the tailpipe. Ansver "b." is more general and, therefore, the more correct of the two selections. RECOMMENDATION: Accept ansvers "a." and "b."
REFERENCE:
Print D-302-605
- NRC RESOLUTION: Concur. Accept answers "a" and "b".
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s .* i t QUESTION: 084 (SRO) 086 (RO) POINTS: 1.00 SELECT the statement that describes the reason (reactor vessel vater level instruments may not be used) for level indication if the dryvell temperature near the reference legs is greater than the RPB Saturation Temperature.
- a. The differential pressur.e transmittars are not environmentally qualified to operate at saturated temperature conditions,
- b. Dryvell pressure is the same as reactor' pressure providing a zero differential pressure (upscale level indication).
+- Actual RPV vater level may be lovt.r than indicated vater level, f
- d. The density of the va er in the reference leg is too lov to provide a usable differential prtssure to measure level.
ANSVER: c. COMMENT: Ansver "c. does not indicate an unusable reading. Throughout the PEI's, if level is not accurate under given conditions, graphs are used for correction factors. Ansver 'c." is a grcas simplification of the problem. Ansver."d." ds more correct because it indicates that the instrument is not usable. RECOMMENDATION: Accept ansvers "c." and "d." Rhc1RENCE: Lesson Plan OT-3034-02-D23-1 NRC RESOLUTION: Concur. Accept answers "c" and "d". l t
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, .- ,o OUESTION: 067 (SRO)
'Ot0 (RO)
POINTS: 1.00 Given the following plant conditions: PLANT CONDITIONS Reactor power 4% Main steam lines are isolated. Suppression pool cooling is operating (both loops). . SELECT _the HAXIMUM Suppression Pool temperature that can_ exist BEFORE Standby 1 Liquid Control (SLC) HUST be injected. FEI-B13, Figure 1, " Boron Injection Initiation Temperature" is provided as Attachment 9.
- a. 121 P.
- b. 120 P.
c.- 110 F. i 109 F. ANSVER: b. CONMENT: The 4% power line is greater than 1205F and less than 122*F; , however 121'T can be the maximum temperature that can cxist ' before SLC must be injected, based on graph interpretation. Graph cannot be differentiated between 120' and 121*F.- Actual temperature appears to be 121.5'F. Therefore, "a." and "b." are correct. l RECOMMENDATION: Accept ansvers "a." and "b."
REFERENCE:
Attached Figure ~1
. NRC. RESOLUTION: Do not concur. Answer "a" is the only acceptabic answer.
Answer Key changed to answer "a". T
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U.aS. NUCLEAR'REGULATORYJCOMMISSION; 9 q' .- ' . REACTOR OPERATOR LICENSE-EXAMINATION 1- ,, v4? REGION 3. l n l'a FACILITY:
- Perry 11& 2, .a REACTOR TYPE: . BWR-GE6 DATE ADi'INISTERED: J90/08/06' ]
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CANDIDATE: INSTRUCTIONS TO CANDIDATE: i 8 l
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Pointryfor.eacniquestionyare indicated in parentheses afterJthe questien.' -. To:- J pass this' examination, you :aust. achieve'an overall grade of at-least 80%. ,
-Excmination. papers will be picked up:four and one half (4 1/2) hours 'a f ter u tho examination starts. ,
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NUMDER TOTALL CANDIDATE'SJ . CANDIDATE'S ' i QUESTIONS .OINTS POINTS OVERALL l GRADE (%)E 94 '98.00 . 1
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.?Allswork~done on.this' examination is my own. I hava neither given bnor: received 5 aid.- ,
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CEndidate's Signature-r i MWe xSTER COPY " n l 1 1
'NRCcRULES AND GUIDELINES FOR LICENSE. EXAMINATIONS ,)
'DuringatheDidministration of this examination >the following-rules apply' ~ !
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!1.; Cheating on;the examination means an'automotic denial of'your' application ;
andLeould result-in more-severe penalties. 1 L'. 2 After the examination has been completed,.you must sign the statement on ! 4 the coverrsheet indicating that the work is your-own and you have-not received or given assistance.in: completing the examinat. ion. This must be fdone after you complete the-examination. l
~3. Restroom-trips are to be limited and only one candidate at a-time may l leave' .
You must avoid all contacts with anyone outside the examination- l room to avoid even the appearance or possibility of cheating.
;4. Une black ink or dark-pencil.only to facilitate legible-reproductions'.
5.oPrint your name in the blank provided in'the upper right-hand corner of the' examination cover sheet. 6.JFill 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' >
noparate sheet of paper. USE ONLY THE PAPER PROVIDED AND DO NOT WRITE ON
'THE BACK SIDE OF'THE PAGE.
H 8;JI.f you write your answers on the examination question page and you need more-space:to-answer a specific-questio- use a separate sheet of the
.pnper provided and--insert it directly at r the specific question. DO NOT WRITELON'THE BACK SIDE OF THE EXAMINATION QUESTION PAGE..
l- 9. Print your name in the uppek right-hand corner of the first page of answer sheets whether.you use the examination question pages or separate sheets cof-paper.- Initial each of the following answer pages. 110L Batore you turn.in your examination, consecutively number each answer
- saeet, including any additional pages inserted when writing your answers on ;the examination quest ion page. =ll.;If you.are.using separate sheets, number each answer and skip at least 3 f
lines between answers to allow space for grading. il2.; 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 m Grror resulting'in'an~inccrrect answer. Write it out. ny , c e
,)i Ik E
* + = =
.e 714.,Tho point valus for.each question is. indicated in parentheses after the-question. The amount of blank space onian 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
-thatris provided may~ count against you. For example, if a question is worth one point and asks for four responses, each of which is worth 0.25 points,.and you-give five responses, each of your responses will be' worth < 0.20 points. If one-of your five responses is incorrect, 0.20 will be deducted and your total credit for that question will be 0.80 instead of 1.00 even though~you got the four correct answers. ;18. If the intent of a question is unclear, ask questions of the examiner-only. )19.nWhen turning in your examination, assemble the cGopleted examination with examination. questions, examination aids and answer sheets. In. addition, turn in all scra; paper.
- 20. To pass the examination, you must achieve an overall grade of 80% or greater.
-21.:There is a time limit of (4'1/2)~ hours for completion of the examination. i L(cr some other time if less thsn the full examination is:taken.)
22- When you are done and have turned in your examination, leave the examin-
, ation area as defined by the examiner. If you are found in this area i while the examination'is still in progress, your license may be' denied or i revoked.
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.REA.f0R OPERATOR. :Pago 14 ' ]
L" e. QUESTION: 001- (1.00)
. SELECT the statement that describes the proper method of verifying the poaition of a valve: after initial positioning of the valve has been completed following maintenance on the valve. The-valve checklist required'
, -pocition is "LX."
- a. Verify it closed by position indication and resistance to turning the handwheel in the closed direction. t b.: Verify it~C .ved by turning the-handwheel 1/8 to 1/4 turn open and reclosing the valve, attach the locking device. ,
c.. Verify-it closed by turning-the handwheel 1/8 to 1/4 turn open and reclosing the valve.
- d. Verify the valve closed by position indication and resistance to turning the handwheel in the closed direction, attach the' locking- .
L device. l l? LQUESTION: 002 (1.00) SELECT the statement-that describes a requirement when performing a system ) l . valve lineup.
- a. . Unaut):orj zed locking devices should be removed when discovered and- '
noted in-the " remarks" section of the: lineup sheet. b.: Missing valve labels should be corrected when discovered by
, -writing the valve name and number.on the valve body.
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- c. A, pencil or highlighter may be used while in-plant to. check steps L done,.the initials blocks may be done.laterlin the control room.
d.- Valves found mispositioned should be repositioned in accordance-with the lineup sheet, unless the Shift Supervisor directs 4 otherwise. -i q. S I l\ . 1 f
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' REACTOR OPERATOR Page 5 :.
7 QUESTION:'003 (2.00) MATCH
-used each ofB.the situations in Column A with the type of tag that would be in Column be identified. Each The MINIMUM tagging required for each situation should not at all. item in Column B may be used once, more than once, or ONLY a SINGLE answer is used for each item in Column A.
COLUMN A COLUMN B (SITUATIONS) ____________ (TYPE OF TAG)
- a. A valve that could'cause 1. Jurisdictional Tag personal injury if _
operated.
- 2. White, Out of Service
- b. A valve undergoing minor ,
maintenance lasting less 3. Red Tag than one shift. 4. Yellow Tag
- c. A red-tagged circuit breaker that could provide 5. MFI Tag power to a grounded bus. 6. LLJED Tag -
- d. A blocked relay in a logic train, 7. ILRT/LLRT Tag-i l
i i, ,
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I t . , i TREACTOR OPERATOR' Paga '6 - {i; LQUESTION:: 0042 (1.00) ('~
. SELECT the statement that describes the quarterly"whole body (WB)l exposure i
l limit for an individual with the exposure history identified >below as per q j l
=PNPP administrative guidelines.
.t l~ I1',XPOSURE HISTORY t
iAge:- 25 years. 1 Lifetime exposure: unknown. l NRC Form 4: not available, exposure history is undocumented. 1
'a. -300 mrem. t
- b. -1000 mrem.
-c. 1250 mrem.
- d. 300'O mrem.
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6 Pagaf 7:
~ REArTORTOPERATOR' s
x: QUESTION: 005' (1.00) SELECT the' statement that: describes the remaining Federal whole body (WB); !
' 'Cxposure'for the current calendar quarter for an individual with the
~
Gxposure history idantified bel' 2 EXPOSURE. HISTORY Age: . 22 years. Lifetime-exposure: 21 rom. Current quarter: 250 mrem. NRC-Form 4: On file, lifetime and current quarter exposure: history is documented.
- a. 250 mrem.
- b. 750 mrem. )
c.- 1000 mrem..
- d. 2750 mrem.
4 LQUESTION: 006 _(1.00) SELECTLthe statement that-describes a circumstance where a radiation work-
- 1parmit;(RWP) is required.
- l J
- a. Area radiation levels are 75 mrem / hour.
th Airborne activity levels are 0.2 MPC due.to noble gas,
- c. Area-neutron radiation levels are 0.25 mrem / hour.
d.- ALjob whers a worker will receive 80 mrem in one shift.- , i t i
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s Pcgs 8; -tEACTOR -. OPERATOR -
.i 7UESTION:jo07 ~(1.00) Hl ~ SELECT the~ statement that describes.a responsibility of any radiation -:workOr.at the Perry plant. ,
LSurveying high radiation areas entered while- on a radiation work
~
Da.
. permit. =
b.. Complying with all of the requirements identified on the radiationi ' [: work permit for a controlled' area to be entered. c.- Evacuating,a controlled area if their self-reading dosimeter 1
. reaches 50% of its maximum reading.
i levels
- d. Refusing to enter controlled areas with general ;0diat on in excess of 150 mrem / hour.
. QUESTION: 008 (1.00) .
SELECT :the' statement that describes the MINIMUM personnel. protective equipment required in additionfast to eye protection speed breaker, when racking out'the
;L1205, Recirculation ~ Pump "B"
- a. Leather work gloves,
- b. 10 KV rubber gloves and leather gloves.
^ ?c. '20 Kv' rubber gloves and leather gloves.
S Dd '. -. 20 Kv-rubber gloves, leather gloves, and 20 KV rubber sleeves. t 9 1 s s-i Y \
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,l'il _Page .9 REACTOR-OPERATOR.. s 0 QUESTION: 009 (1.00) SELECT the plant staff i nember who should not be a member.of.the fire brigade team.
- a. Supervising $perator at the controls. .
- b. Non-licensed operator.
- c. Rad-waste operator.
- d. . Electrical maintenance supervisor.
' QUESTION: 010 (1.00)
SELECT therstatementethat describes a condition that must be met for the-
-, Supervising Operatoriat the controls to be. temporarily relieved for personal reasons..
- a. 'The: Shift Super.isor's approval must be obtained, b.. A verbal turnover of status of work in progress must be completed.;
c .? The "So Relief / Turnover Checklist" must be signed by the oncoming operator,
- d. The oncoming = operator signs into the Unit-Log as. Supervising
_ Operator at the controls. n (d: , 3 1 t
n .. .. .. . . u Page 10'
- EACTOR OPERATOR; o
4 i
'QUESYION: 011 -(1.00)
, 1 Given the fo11'owing - surveillance data:
~
SURVEILLANCE DATA' 3;
' Recirculation pumps are in. fast speed.
1B33-R613: 45 M1hm/hr. Jet-purip-data (Mlb1/hr):- 7 1033-R611A, JPS: 1.6
- 1B3 3-R611B, . J.P10 : 1.7.
'1B3 3-R611C,- JP15: 1.7 1B33-R611D, .JP20: 1.6 APRM flow readings 5'(%):
Div.-1:1 A: 39 E: 39
.Div. 2: B: 36 F: 36 Div. 3: C: 40 G: 41 Div. 4:- D:.38 H: 38 SELECT the statement that describes:the'results of the performance of SVI-C51-T0026, "APRM Flow Biased Power-Flow Verification.." Copies;of A003,
" Percent Drive = Flor vs. Total Ccre Flow" and SVI-C51-T0026 are provided.as:
' Attachment'1.
a '. - 'Both 'div. 1 and both div. 3 APRMs are-inoperable, b.- The' G" APRM is inoperable.
;c.- Both:div. 3 APRMs arelinoperable.
- d. Surveillance is satisfactory, tPJ ,o
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i. R 36i > l REACTOR. OPERATOR- Paga 1' o J F l QUESTION: 012- ;(1.00) SELECT the'stStement that identifies the significance of_a," YELLOW"
. indication on the Emergency _ Response Information Systemc(ERIS).
j7 a. ' Requires imme'iate d operat'or. action. h
~
~ b.. It is:in a~"NOT COMMANDED" state.
- c. It is approaching its parameter limit.
- d. It.is normal, satisfactory conditions exist.
. QUESTION: 013. (1.00)
= SELECT the statement that describes the response of the Control' Rod Drive-
.rated' Hydraulic system to a loss of service / instrument air while operating =at-conditions.
a.. Flow' control _ valve (FCV): Fails as is-Scram inlet and outlet valves: Fail open-Scram l discharge volume (SDV)> vent and-drain valves: Fall.open. s b. Flow. control valve-(FCV):
. Scram. inlet andLoutlet. valves: Fails' closed Fall- ~ closed Scram 91scharge volume (SDV)' vent and drain valves:
Fail;open i
- c. Flow control valve (FCV):
. Scram inlet'and outlet valves: Fails closed-
+ Fail open i
. Scram discharge volune-(SDV) vent and drain valves:- Fail' closed. l
- d. . Fl'ow control valve (FCV): 'l scram inlet and outlet; valves:. .Fa'ils open i Scram: discharge. volume 1(SDV), vent and drain valves:
_ Fail.open j Fail Eclosed 'l l
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1 1
REACTOR 10PERATOR7 ' Pa g e' _12 .'
-QUESTION:!.014 ( 1. 0 0 ) --
u+ SELECT the. statement.that describes the operation 1of-the-Redundant
-Reactivity: Control System's (RRCS)-Alternate Rod Insertion-(ARI)' valves whenLARI= actuates,
- a. Energize'from"the'125 vde bus EDIA (1R42-SO12).
- b. Deenergized from the 125 vdc bus-EDIA (1R42-S012).
e.- Energize;from the Reactor Protection System (RPS) bus B.
- d. Deenergize from_the Reactor Protection System-(RPS) bus B.
QUESTION: 015 (1.00) SELECT the statement-that describes the indications that verify,a control-rod-is coupled toJits1 drive mechanism when it.is notched out from; position-
- (3. .
- a. " FULL OUT" LED remains lit, " ROD OVERTRAVEL'? annunc'iator does not calarm.
- b. " FULL OUT" LED goes ' out, " ROD OVERTRAVEL" annunciator: alarms'and immediately clears.
~
- c. -Rod position indication goes blank then indicates "48" again,;" ROD OVERTRAVEL" annunciator alarms and immediately clears.
- d. Rod position indication goes blank, " ROD _ OVERTRAVEL" annunc'iatcr alarms.
l 4.
f I 1 REACTOR OPERATOR Pags 13:
- QUESTION: 016 :
(1.00). Given(the-folloving plant conditivas: p PLANT: CONDITIONS-Reactor power:' 45% , Generator load: 480.MWe
-Control rods-in' group.9 are being withdrawn by' notch withdrawal'from notch 127to. notch 24 (the desired rod pattern.line).
SELECT the statement that describes a result of the main turbineLfirst stage shell pressure input to the Rod control and Information System' .
- (RC&IS)Lfailing to its maximum pressure value.
L , a .- Any additional rod withdrawal is limited to four notchestof. > continuous. travel for any selected rod in the. allowed-sequence. s
- b. Any-further rod withdrawal is limited to'two actches of. continuous-
. travel for'any' selected rod in the allowed' sequence,
- c. Any;further rod withdrawal is NOT limited by notch withdrawal' L constraints in-the Ellowed sequence.
.d. Any further rod withdrawalfof'any rod is blocked.
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,j Paga 14 lif[kACTORIOPERATOR
+ .
- [.h h' njQUESTION:LO17; (1 00)
. 4 ty
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CGiven the following plant conditions:
~
JPIhNT' CONDITIONS l e [ Reactor power: 26% '
. Control rod withdrawal: In progress.
T'
- Bypass valve 1: Open.
j' J Load 1Amiter setpoint:. 120 MWe. vr >
> iSELECT-the statement that describes the potentia 1' consequences of this. I g joituation occurring during.
li
- a. The rod control and information system (RC&IS) may allow non-conservative control rod withdrawals.
o -
-;c b.. :The' main condenser may lose vacuum due to the direct addition of.
steamtto' main condenser. } f l f c. The. reduc'ed' extraction steam; flow from the main ~turbinet may cause. excessive erosion of=the low pressure turbine. L .d. lThe reduced feedwater heating due to the-bypass valve opening - [
-could>cause-fuelLthermal limits to be exceeded.
a l' ,, l ' ! zQUESTION: 1018 L(1.00) p SELECT-theareason a Recirculation Pump must' operate for 15. minutes after it
- , ;ic Otartedifrom rated conditions before a second start attempt is ,
b :parmitted. b ca. LTo ensure motor bearings have returned to normal temperatures., li ..
- b. 'To. ensure.the;pumpLimpeller has cooled after rubbing the. casing.
'c . To' ensure the. motor windings have cooled to rated temperatures., }
d.- To ensure CRD-flow has-flushed the seal of any impurities. b . L [ l. l ;c
.f - Pago 15
' REACTOR. OPERATOR I
.r i
TQUESTION: 019- (1.00)
- SELECT:the statement that describes the effect on the Recirculation system for anLactuation of the end of core life (EOC) interlock following a mala turbine trip.
- a. At lesr than 40% rated power,-the low frequency motor' generator (LFMG) and the CB3 and CB4 breakers will-trip for each pump- .
7 l . b. At greater than 40% rated power, the CBS breaker will trip, the low frequency motor generator (LFMG) will start, and.the CB2 t breakers will close for each pump.
- c. At greater than 40% rated poker, the CB3 and CB4 breakers will trip, the los frequency motor generator (LFMG) will' start,1 and the CB2 breaker will close for each pump,
- d. At less than 40% rated power,=the CBS breaker will trip, the low '
frequency motor. generator (LFMG) will start, and the.CB2 breaker. will close for each pump. l i 4 s > 1 4-b- E' s I
O 1
> REACTOR 1 OPERATOR Paga 16~
i J TQ UES'110N:_020- (2.00) l MATCH the-recirculation system response in Column B to each of the ' opbrating conditions listed in Column T. The recirculation system is-ini.tially operating at rated core flow. Column B responses may be used ,
' onc'@ , more than once,-or not at all. ONLY a SINGLE answer is used for'each. L item in Column A.-
COLUMN A COLUMN B
.(OPERATING CONDITIONS) (RL'CIRCUL\ TION SYSTEM RESPONSE) a..The "A" circulating water. 1. No effect on the recirculation pump trips and condenser system. '
pressure increases to 5.5 inches of Hg absolute. 2. Flow control valve (FCV) is . limited to 48% maximum flow. i
- b. Loop flow controller _
Flow control valve (FCV) is output signal. fails high. 3 .' motion is-inhibited =(no'
- c. RPV water level is +206 further motion allowed),
inches and only the "B" RFP is in service. 4. The recirculation pumps downshift to lowispeed.
- d. HPU tank oil level is empty. 5. Tr.e low' frequency motor.
generator (LFMG) breaker trips. i i ; r s i
cREACTOR OPERATOR pag 3 17 <
- QUESTION: 021 (1.00)
' SELECT.the statement that describes a situation when the Low Pressure-Coolant Injection (LPCI) From RHR A Shur-off-Valve (1E12-F042A) will'open. '
- a. A LPCI signal'is present and the downstream loop pressure is GREATER THAN 530 psig.
- b. A LPCI signal _is present and the RHR A injection valve (lE12-F027A) is open.
- c.
The control switch is taken to open and the blue light-is ' illuminated indicating downstream loop pressure is LESS THAN 530 psig, d. The control switch is taken to open and the RHR A injection valve 1 (1E12-F027A) is open. ,; QUESTION: 022 (1. 00) SELECT.the statement that describes the potential effect'on the."B"<and Rnoldual Heat Removal (RHR) subsystems as indicated by the receipt of the"C" "RHR B OUT OF SERVICE" and "RHR C OUT OF SERVICE" annunciators:due to the
'woterleg motor pump (lE12-C003). supply breaker tripping.
-a. The-pumps are in danger of electrical overloaduin the event of a.
LOCA' initiation and must be declared inoperable. .
- b. The pumps may not inject to the reactor; vessel within their analyzed injection time in the event sf a LOCA.
c.. The pumps will not str.rt in the event of a -LOCA since their supply breakers are interlocked open.
- d. The pumps may'not achieve design flow due to excessive pump.
cavitation. ,
e REACTOR OPERATOR ~ -PageE18 I iQ'UESTION: '023 (2.00)
- MATCH-the reactor water cleanup-(RWCU)-system response inl Column B.to'each-of the operating conditions in Column A. .The RWCU system is initially in-its normal full-power configuration. The responses in Column B may be used once, more than-once, or not at all.
item in Column A. ONLY a SINGLE answer is used for-each' COLUMN A COLUMN B (OPERATING CONDITIONS) (RWCU SYSTEM RESPONSE)
- a. RPV water level is 1. - Only the pump suction inboard
+104 inches. isolation valve (1G33-F001) isolates.
- b. RWCU-differential flow is 60 gpm. 2. Only the pump suction outboard
- c. The keylock switch for the isolation valve (1G33-F004).
isolates.
"A" Standby Liquid Control pump.is taken to "ON". 3. Both the pump suction inboard
- d. RWCU non-regenerative heat and outboard isolation-valves exchanger (liRHX)' outlet (1G33-F001 and -F004) isolate.
temperature is 132 F. 4. No isolation action occurs.
}
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'tREACTOR' OPERATOR
( Page 19' iQUESTIONi;024 - (1. 0 0) -
'!Given'the:following initial ~ plant' conditions:
INITIAL PLANT CONDITIONS RHR loop "B" 'is ,in Shutdown Cooling (SDC) mode. Coolant-temperature is 300 F. RPV_ pressure is 65 psig.
'_ SELECT theEstatement that describes the effect on the Shutdown C:'"r ' ~'
N -(SDC) 1E12-F008)Suction if Isolation the 4.1d ~KV Inboard and Outboard. bus EH12 trips, Valves (1E12-F009 and a. lE12-F008 and IE12-F009 will not isolate at 135 psig due to;a' loss of power. b. 1E12-F008 and 1E12-F009'will-both isolate when pressure reaches 135-psig.
- c. 1E12-F009 isolaths when_ pressure reaches 135 psig,i1E12-F008 wills
-not isolate due:to a-loss of power.
H
' d.
1E12-F008 isolates when pressure reaches 135 psig, 1E12-F009 will; 4 not isolate due to a loss: cf power. t q:= K, . I w b i 'F
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"[REACTOROPERATOR:
~ Paga 20-4 4
; QUESTION: 102 Si ~- (1. 00) m ei-f GIVen:the fol' lowing. plant conditions:
PLANT CONDITIONS-RPV water level: +16 inches for.105 seconds. iDrywell-pressure: . .2.0 psig
' Residual Heat-Removal (RHR) Pumps: Not running.;
SELECT the. statement that describes the effect that the Low Pressure Core Spray 1(LPOS) pump motor-breaker failing to close'would have on the-(Automatic Depr3ssurization System (ADS). .
-l a.- ADS automatically _. actuates at this time. !
- b. ADS will NOT automatically actuate.
- c. ADS will actuate 105 seconds after the ADS "A" manual initiation pushbuttons are depressed, p d. ADS willjactuate when the ADS "B" manual initiation pr.shbuttons
, . are depressed,
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~ REACTOR / OPERATOR Pagb21.
.l t
QUESTION:'026' ( 2.00) MATCH each:of;the Reactor. Protection System (RPS) A'with its Technical 1 Specifications basis in column Bitrip parameters in column
.mnyrbe used once, more than once, or not at all. ONLY aItems in column B uncd for each item in Column A. SINGLE answer is COLUMN A COLUMN B (RPS TRIP. PARAMETER)
..___..___________-- (TECHNICAL SPECIFICATION BASES)J
____ .__ ...... _______________ - 1 a.' Main steam line high 1. Anticipates a pressure and radiation. neutron flux increase and provides pressure and fuel
- b. Main steam line isolation thermal / hydraulic safety limit valve clcsure. protection.
- c. Reactor vessel high water
-level,
- 2. Compensates for reactivity ,
addition and. subsequent 1 power
- d. Reactor vessel steam dome riso addition, due to cold water high. pressure.
3. Reduces!the continued failure of f..el cladding. 4.- Compensates for reactivity addition from void collapse ' and resulting power rise.
.5. Provides protection from exceeding the low pressure /
low flow thermal power. safety limit. 4 {. 5
't 7 . I 3
" REACTOR; OPERATOR .
4 Page 32 h l QUESTION: 027 (1.00) SELECT the condition that would cause-a reactor scram,
- a. Turbine Stop Valves (TSVs) "A" and "B" are closed and turbine
, first stage shell pressure is 50% power. '
- b. Turbine Control Valves (TCVs) "A" and "C" emergency trip supply (ETS) pressure is 650 psig and first stage shell pressure _is 25%
power. fc. Turbine Control Valves (TCVs) "A" and "D" emergency trip. supply (ETS) pressuru is 230 psig and first stage shell-pressure is 45% power.
- d. Turbine-Stop Valves (TSVs) "B" and "C" first stage shell pressure is 40% power.are closed and turbine QUESTION: 028 (1.00; SELECT the effect that a loss of one of the two position probes for a control rod would have on the Rod Control and Information System (RC&IS).
- a. RC&IS generates a select block for that rod, the block can be cleared by bypassing the rod at the rod bypass file in each of the Rod Action Control System (RACS) cabinets.
- b. RC&IS generates an incert block for that-rod, the block can-be.
cleared by cross-conne_ ting the probes with jumpers 11n the Rod Position Multiplexer cabinets, c. RC&IS generates a withdraw block for that rod, the block can be cleared by substituting rod position information from the operable probe.
- d. RC&IS prevents any further gang movement of that "od, to move the rod individually, the rod must be bypassed at the rod bypass file in its associated Rod Action Control System (RACS).
II, l'
REACTOR .. OPERATOR ; Pcga 23
-QdESTION: '029 (1. 00)
' SELECT the' statement that describes a condition in which the Intermediate-R0nge' Monitoria.g (IRM) circuitry is NOT functioning correctly.. The reactor mode switch is in "STARTUP/ HOT STANDBY. "
N-33 ' a' . IRM "A" indicates 90/125 scale on range 7 and no rod block or il scram signal is present. fi b . -- IRM "B" indicates 122/125 scale on range 6',
. present,-and.IRM upscale trip light is lit.
a rod _ block is
- c. IRMs "D" and "A" both fail full upscale and a reactor scram-occurs.
- d. IRM "C" mode / test switch is placed to " STANDBY".and a rod block and half-scram' occurs. j i
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; REACTOR OPERATOR- '
s Paga;24; R I s l m
; QUESTION: 030-'
(1.00) C -Givenithe:following! plant' conditions:
~ PLANT CONDITIONS i
.The' reactor. node' switch position: "STARTUP/ HOT STANDBY."-
Reactor power'is midscale on rang 6 of=the Intermediate Range 1
- Monitors (IRM). I Source Range Monitor (SRM) B-is failed upscale'and is bypassed.
All'IRMs are selected to range-4. I
- SELICT the-statement that describes an operating condition that will?cause tha' Rod-Control =and Information System (RC&IS) to generate a rod block.
a.
-The SRM channel'A indicates 7 E.4 cps and its detector is' fully inserted..
m ' b. : The SRM channel C indicates 50 cps and its-detector-is partially withdrawn.- i
'c.
The:IRM inserted. channel B-indicates 30/125 scale and its detector.is fully- I d. The fullyIRM channel F indicates 8/125 scale and its detector is not inserted. i-l l l. I ll l 1 t s l; l, I
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+ , ,
Pag 3 25 REACTORf0PERATOR' IQUESTION:-031 (1.00)- SELECT thefstatement that describes [the Average Power Range Monitoring ;
;cystemL(APRM) input 1to the recirculation ~ flow control ~s'ystem.-
- a. hPRM~"E" output is compared with actual core flow to generate,the- !
loop flow demand signal- ;
- b. APRM' " A" output is compared - to actual core flow signal- to shif t '
the loop flow controller to manual if not in the allowed area-of no the1 power-flow operating-map. i
- c. APRM "A" output is compared to the load-following circuit's load demand error signal to generate the flux error signal, matching 1 reactor power to generator load.
- d. APRM "E" output is compared to the output of the flux controll'er to generate the flow demand signal, a
i (QUESTION: 032 (1.00)
~
SELECT-the statement that describes an operational condition when the Rstctor Core Isolation Cooling (RCIC) system would FAIL to initiate.~
~
+
1
- a. RPV water level-is +18 inches and the B21-N692A level' instrument- ,
is-failed upscale. i b .- RPV water level is +20 inches and the 125 vdc bus ED1B is l deenergized. ,
-c. RPV water level is +105 inches and the-125 vdc bus EDIA is .
.deenergized. :
-d. RPV. water level is-+110 inches and'the reference-leg for the .
B21-N692A and B21-N692E level instruments is ruptured. e i t'
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[ REACTOR OPERATOR Page'26 5
~ QUESTION: 033 (1.00)
SELECT the statement that describes the sequence of actions necessary to racet and'open the Reactor Core Isolation Cooling-(RCIC) turbine trip and throttle (T&T) valve after it trips due to overspeed protection.-
- a. .
Take the'T&T valve control switch to close until the green light is on and the red light is off, reopen the T&T valve with the control switch, reset the mechanical linkage locally, b. Reset the T&T valve mechanical linkage locally, take the T&T-valve control switch to close until the green light is on and the red light is off, reopen the T&T valve with the control switch, c. Take the T&T_ valve control switch to close until the green light is on and the red light is off, reopen the T&T valve with the cantrol switch. d. Take the T&T valve control switch to close until the green light is on and the red light is off, reset the T&T valve mechanical linkage locally, reopen the T&T valve with the control switch. -QUESTION: 034 (1.00) SELECT the statement that describes operation of the automatic pump suction cource transfer of the Reactor Core Isolation Cooling (RCIC) system. a. The CST suction valve (1E51-F010; opens and the suppression pool (SP) suction valve (1E51-F031) closes on a SP low level signal. b. The supp~ession pool (SP) suction valve (1E51-F031) opens and the CST suction-valve p.E51-F010) will close on a SP high level signal. c. The suppression pool suction (SP) valve (1E51-F031) opens and the CST suction valve (1E51-F010) closes on a CST high level signal, d. The CST suction valve (1E51-F010) opens and the suppression pool
.(SP) suction valve (1E51-F031) closes on CST low level signal.
I, l
u i REACTORf0PERATOR - Paga 27 QUESTION: 035 (1.00) SELECT the statement that describes the Automatic Depressurization System (ADS) interrelationship to the DC electrical system.-
- a. All 8 ADS valves will open if BOTH-subchannels in EITHER the A channel logic powered from division I 125 vdc OR.the B channel
. logic powered from division II 125 vdc are tripped.
b. All 8 ADS valves will open if ONE of the subchannels in EITHER the
- A channel logic powered from division I 125 vde OR the B, channel logic powered from division II 125 vde are tripped.
c. Only 4 ADS valves will open if division II<125 vdc power is ' available and ONE of the subchannels in the B channel logic is tripped. d. Only 4 ADS valves will open if division I 125 vde power is ' available and BOTH of the subchannels in the A channel logic are
-tripped.
i
+ REACTOR OPERATOR Paga 28 7 f
a QUESTION: 036 (1.00) i Given the'following plant conditions: i PLANT CONDITIONS
- Low Pressure Coolant Injection (LPCI) signal is present.
Residual' Heat Removal its heat exchanger. (RHR)' loop A is injecting in LPCI mode through The heat exchanger bypass valve (1E12-F048A),is closed.
- uthe Residual Heat RemovalSELECT the statement (RHR) loop "A" that describes the condition that must oc 3
to suppression pool cooling. mode. a. RHR can be realigned after the LPCI initiation signal clears.
- b. ftAli9 erb RHR can beVunder the present conditions.
c. RHR can using be. realigned the reset pushbutton.after the LPCI initiation signal is reset- .3' f d.- RHR can be realigned by installing jumpers to override the LPCI signal.
- QUESTION: 037- (1.00)
SELECT the statement that describes the operation of the Residual Heat Rtmoval (RHR) ! ! (LPCI) signal occurs. system ten minutes after a Low Pressure Coolant Injection : 1
- a. .
.When both received, the loop
.:RHR containment AND drywell high pressure signals are.
"A" will . realign and automatically spray
)- containment. .he b. When either the containment loops "A" OR drywell high_ pressure signals are received, both RHR - and "B" will realign and automatically spray the containment, L
- c. RHR loop "A" will realign and automatically spray the containment.
- d. Both T.HR loop "A" and "B" l containment. will realign and automatically spray the 1
L
' REACTOR OPERATOR PCga 29-j
.QUESTION:~038 (1.00)
. SELECT the immediate action required.if the fuel pool cooling and cleanup (FPCC) return 'line to .the fuel pool ruptures causing the receipt of the
-following annunciators.
Puel is being offloaded from-the reactor. ANNUNCIATORS- , CONTAINMENT FUEL STRG POOL LEVEL LO LO CONTAINMENT FUEL STRG' POOL LEVEL LOW i FPCC SURGE TANK A LEVEL HI/LO' FPCC SURGE TANK B LEVEL' HI/LO a. Irrespective of location of the refueling platform, lower any loaded bundles until the slack cable light illuminates. b. Movo any bundles located in the inclined fuel transfer machine and-place them in a fuel storage rack in the spent fuel pool, c. Immediately stop all fuel movement and evacuate the containment and drywell. d. Lower any loaded fuel bundle into any vacant fuel cell in the reactor vessel. QUESTION: 039 (1.00) SELECT the statement that describes a situation where ALL power would be interrupted to the mainhoist. fuel hoist on the refueling platform preventing ANY1 ifurther motion of the . a.- A fuel bundle is raised until the overhoist light is illuminated.
-b.
A fuel bundle is1 raised until the grapple normal up light is illuminated, c. The refuel platform-the traverse is inpushbutton. system stop tr.c predetermined " red zone" lighting
- d. Asfuel bundle is being removed from.the inclined fuel transfer.
L carriage while a control rod is being withdrawn. l t
i i,; ' P go130 ! I ' REACTOR' OPERATOR' H J-Y ! QUESTION: -040 : (1. 00)L LGiven the following plant _ conditions: ,
TheLturbine control valves (TCVs) throttle closed, the bypass valves (BPVs) throttle-open to compensate, once-the BPVs are fully open reactor pressure increases causing.a scram on high pressure-or neutron flux.
- d. .The turbine control' valves (TCVs) throttle closed, the bypass j
valves-(BPVs) throttle open, reactor pressure remains fairly j constant, reactor power-increases slightly due to_ reduced D feedwater-heating.
'l i
a 1 4 I 1 s
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REACTOROhkk5 TOR Pagi 34- ; t i
, . QUESTION:-045- (1.00)
Givan theLfollowing plant conditions: 1 5 l PLANT CONDITIONS Reactor power: 100%. I Generator load: 1245 MWe. Standby pressure regulator: In " TEST."' 1 SELECT the action required if the Electro-Hydraulic control (EHC) turbine 1 inlGt pressure sensor fails HIGH for the following plant conditions. t. 1 a.- Control reactor pressure by using the bypass valve jack and commence a power-reduction,
- b. Match steam flow to reactor power by reducing the maximun combined flow setpoint,
- c. Manually scram the reactor and maintain cooldown rate less 'than:
100F/hr by' closing.the MSIVs. d~. Manually. scram the reactor and manually control reactor pressure below'1065 psig~using the bypass valve jack. , _ QUESTION: 046 (1.00) . 3 LSELECT-the statement that describes the relationship between1the condensate
~
System and the Control Rod Drive Hydraulic (CRDH) System.
~
l
- a. CRDH normal water supply is from the condensate booster pump (CBP) suction, its alternate supply is.from the-Condensate Storage Tank (CST). ,
t b.- CRDH normal water supply is from the hotwell pump suction line, ilts alternate supply is from the Condensate Storage Tank (CST). ,
- c. CRDH normal water supply is from the hot surge tank, its: alternate i supply is from the demineralized water header.
- d. CRDH normal water supply is from the hot surge. tank, its alternate supply is from the condensate Storage Tank (CST). }
'O at tREACTOR OPERATOR Paga 35 '
i i 1
-QUESTION::047 -(1.00) t SELECT the statement that describes the operator action required when the i Hot SurgeinTank resulting level controller the.following plant _ malfunctions conditions. causing its level to increase PLANT CONDITIOtlS-
" HOT, SURGE TANK' LEVEL HI" annunciator alarms.
Hot surge tank level is off-scale high. ! i
- a. Immediately reduce core flov to minimum, b
- b. Manually open the reactor feed booster pump minimum flow valve. l
- c. Immediately trip the main turbine, d.
Trip all hotwell pumps and condensate booster pumps. ! QUESTION: 048 (1.00) t SELECT the operational conditions required that~will' automatically start
' the Motor-driven Feedwater Pump (MFP) when'the "A"-Reactor Feedwater Pump Turbine (RFPT) trips, leaving only the "B" RFPT in service.
a.- MFP control switch position: AUTO.
"A":RFP hydraulic trip header pressure:. 90Epsig.
MFP suction valve position: Closed.
- b. MFP control switch position: OFF.
-"A"_RFP hydraulic trip header pressure: 25 psig.
MFP suction valve position: Open.
- c. MFP control switch position: OFF "A" RFP hydraulic trip header pressure:
65 psig MFP suction valve position: Closed.
- d. MFP control switch position: ; AUTO "A" RFP hydraulic trip header pressure:
65 psig MFP suction. valve position: Open, i
) J cREACTOR-OPERATOR Paga 36 !
i
; QUESTION: 049' -(1.00) 1
-SELECT,the signal.that provides an interlock to' ensure adequate. net-
- po31tive suction head.for the components of the recirculation system,
- a. ' Differential temperature between the recirculation loop suctions. I
- b. Hydraulic Powe.- Unit pump discharge pressure,
- c. . Differentia
- temperature between the Recirculation Pump suction.
and reacter steam dome.
- d. Reactor water level (level 4), f e
QUESTION: 050 (1.00) W..a Given the following initial plant conditions: PLANT' CONDITIONS-i Reactor power: 65% Reactor Feedwater Pump Turbine (RFPT) "A" and "B" controllers are in automatic. S' ELECT the state. ment that describes the plant response to the Main Steam: Line. (MSL) "A" flow detector failing downscale,
- a. Reactor water level Will increase initially, then return to'the tape-set value,
- b. Reactor water level will decrease initially, then return to the tape-set value. <
c.
.level Reactor water level will increase until it stabilizes at some higher than the tape-set "a ue.
- d. Reactor water level will decrease until it stabilizes at some level lower than the tape-set value.
A }
- REACTOR OPERATORi ,
PagaL37 j I 3 4
, QUESTION: 051 (l'. 00 );
a
-Given the following initial plant conditions:
PLANT CONDITIONS. ' 13.8 KV Bus L10 is~ powered from startup transformer 200-PY-B.. ' HPCS diesel generator is tagged out of service for maintenance. 1
' SELECT the statement that describes the expected response of the AC electrical distribution system following a main turbine trip due-to a main 1 gtnerator Above" differential is provided current lockout as Attachment 6. trip. Figure R10-2A, "4.16 KV and- !
a .~ The Lil'and bus,Ebut canL12 be buses manually willtransferred, NOT automatically transfer to the1L10 b.
.The L11 and L12 buses will automatically _ transfer to the:L10 bus.-
c.. TheJEH13 bus will be deenergized since the HPCS diesel' generator:- is tagged out.. ! j d. Bus L12-will powered fromNOT transfer startup to the200 transformer L10PYB. bus when the Llo bus is ' 1 i I s i j q
REACTOR OPERATOR ', Pago 38
. QUESTION: 052 (1.00)
SELECT the condition that will cause the static transfer' switch in the-PlGnt Vital Balance of Plant uninterruptible power supply-(BOP-UPS) to automatically shift. system
- a. Low voltage sensed at the output of the BOP-UPS inverter will transfer the BOP-UPS to a bypass transformer powered from bus ,
EF-1-D.
- b. High voltage sensed at the output of the BOP-UPS inverter will transfer'the BOP-UPS to regulating transformer FB-1-R. .
c. A failure of battery 1A's normal and reserve battery chargers for more than 15 minutes will transfer the BOP-UPS to regulating transformer FB-1-R. H
- d. A ground fault l sensed on the BOP-UPS bus V-1-A will transfer-the BOP-UPS to a bypass transformer powered from bus EF-1-D.'
\
' QUESTION: 053 (1.00) <
Given the following plant conditions: PLANT CONDITIONS Annunciators DC BUS D-1-B UNDERVOLTAGE BATTERY 1B DC SYSTEM TROUBLE
' Reactor water level is increasing.
Main steam-line-flow "D" is downscale. , LSELECT the operator action required for these plant conditions.
- a. Select (or verify selected) reactor narrow range level channel B.
. b. Select (or verify selected) reactor narrow range level channel A.
- c. Trip the main turbine-at the front standard,
- d. Locally trip the "A" Reactor Feedwater Pump Turbine (RFPT).
4 Pago 39
' REACTORt OPERATOR-j 1
~
QUESTION: 054- (1.00) l 1
' SELECT the> statement that describes an operational-condition that will i
' ccuSe.the division III Diesel Generator-(D/G) to" automatically trip while. ,
1 it:is operating.. i
- a. RPV water level: +10 inches..
Drywell pressure: 2.9 psig. , Engine' jacket water temperature: 215 F. t
- b. RPVJwater level: +165 inches.
t Drywell pressure: 1.5 psig. Engine crankcase pressure: 5 psig. ,
- c. RPV water level: +165 inches.
Drywell pressure: 2 psig. Engine bearing oil pressure: 14 psig,
- d. RPV water level: +32 inches.
Drywell pressure: 0.5 psig.
- Engine speed: 450 rpm.
.t o ' QUESTION: 055 (1.00) '
1
-SELECT-the statement that describes the correct Diesel Generator (D/G) control' manipulation when it'is being operated in parallel with the grid..
i a .1 Generator VARs are increased to adjust the power factor by taking i
> the D/G voltage regulator switch to " RAISE."
- r. b. Generator VARs are decreased to adjust' power factoriby taking-the :
4 [ D/G governor' switch to " LOWER." p L c. Generator kW is increased-to prevent a reverse power trip by L taking the D/G voltage regulator switch to " INCREASE." I .d.. Generator-kW are. decreased to prevent overloading the diesel by taking the'D/G governor switch to " RAISE." i l l' l l}
. , , , r. - '- - -
J 3)
' o - - s Pcg9/40: (
7 REACTOR. OPERATOR . J b . LQUESTION:',056 (1.00) ; i y i Giyentthe following initial ~Offgas system' alignment: ] INITIAL ~OFFGAS LINEUP ; Charcoal'adsorber control switch: " BYPASS". j Charcoal adsorber bypass valve (1N64-F045): Open.
-Adsorber. inlet and outlet valves Closed, j (IN64-F051A-D and 1N64-F053A/B): -
"A" Steam Jet Air Ejector (SJAE): In service.
SELECT the expected realignment of the offgas' system valvor when the following. radiation monitor indications occur. , RADIATION MONITOR INDICATIONS 1'
-The K601A post-treat radiation monitor is alarming and its "HI"
' indicator light is illuminated.
The K601B post-treat radiation monitor is alarming and its "DOWNSCALE/INOP" indicator light is illuminated, y i
- a. Charcoal adsorber bypass valve (1N64-F045): Open. t Adsorber inlet and-outlet valves: Closed. q p (IN64-F051A-D and 1N64-F053A/B) .
L Offgas' discharge valve (1N64-F632): Open. C E L b. Charcoal adsorber bypass valve (IN64-F045): Closed. Adsorber inlet.and outlet valves:' Open. (1N64-F051A-D and 1N64-F053A/B) Offgas discharge valve (1N64-F632): Open. j c.- Charcoal adsorber bypass valve (1N64-F045): Closed. l Adsorber. inlet'and outlet valves: Open.- (1N64-F051A-D and 1N64-F053A/B) Offgas discharge valve (1N64-F632): . Closed.
- d. Charcoal adsorber bypass valve (1N64-F045): Open.
Adsorber inlet and outlet valves: Closed. (1N64-F051A-D and 1N64-F053A/B) a Offgas discharge-valve (IN64-F632): Closed, r
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; REACTOR OPERATOR- Pago 41 'l QUESTION:=057 (1.00) '
SELECT the condition of the containment ventilation exhaust radiation monitors (K609A-D) that will cause an isolation of the containment' vent and:
~
purgeisystem inboard valves. a'. K609A is alarmitig "HIGH-HIGH" and K609C is alarming "INOP. "
- y. b.- K609B is alarming "INOP" and K609C is alarming "HIGH-HIGH."
- c. K609D is alarming "DOWNSCALE" and K609C is alarming "HIGH-HIGH."
- d. K609A is alarming "HIGH" and K609D are' alarming "DOWNSCALE." '
QUESTION: 058 (1.00) SELECT the statement that describes the response of the Fire Service-Water ! cyctem pumps to an inadvertent actuation of the deluge system for the main transformer,
- a. The diesel fire pump starts when pressure is less than 120 psig ,
for 20 seconds, and the electric fire pump starts at 105 psig in the fire header,
- b. The electric fire pump,is started by the receipt of the deluge
' initiation signal, the diesel fire pump starts when pressure is
'less than 105 psig for 10 seconds in the fire header.
- c. The electric and diesel fire pumps both start when the deluge initiation signal is received.
- d. The electric fire pump starts.at 120 psig, the_ diesel' fire pump starts when pressure is less than 105 psig for 10 seconds in the fire header. ,
l (L
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'Pags,42-h - REACTORf0PERATOR- ,
f.
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.i QUESTION 0.59 - (1. 0 0) -
~ Giv0n theffollowing plant conditions:.
-PLANT CONDITIONS A loss 'of all of fsite power (LOOP) . has occurred. ,
All Diesel G aerators ~ (D/Gs) are' running and tied to their buses.- r
' SELECT the statement that describes the expected alignment'of the: Control ,
, Room Heating, Ventilation, and-Cooling (HVAC) system. -Figure M25/26-1C,
" Control' Room HVAC and Emergency l Recirculation Systems" is provided:as
' Attachment 7.
- a. The supply fan and the return fan are running. ll
- b. The supply fan is running and the return fan is stopped. !
- c. The supply fan and the return fan are stopped..
4 d.: The supply fan is stopped and the return fan is running.. 4 t
, , +
- REACTOR OPERATOR Pcgi 43 QUESTI N: 060- (1.00)
SELECT the operator action required to be.taken by procedure if the "A"- racirculation pump trips while operating at rated power.- Plant conditions-following the recirculation pump trip are listed below. Figure'A006,
" Power to Flow Operating Map," and ONI-B33-2, Attachment-1, " Thermal Power x
J v$rsus Core Flow" are provided as Attachment 8. i i i PLANT CONDITIONS AFTER THE "A" RECIRCULATION PUMP. TRIPS Reactor power:
-Core flow:
45% with 6% bandwidth oscillations. 42 M1bm/ hour
- a. Arm and depress the RPS manual scram pushbuttons.
- b. No immediate operator actions are required.
c. Insert control below rods the limits in reverse sequence to lower reactor power of ONI-B33-2, Attachment 1. ' d. Insert the CRAM rods to lower reactor power below the limits of ONI-B33-2, Attachment 1. I i L l'
REACTOR OPERXTOR' 1Pcg3 44: , 1 QUESTION: 061 (1. 00)' Given the following initial plant conditions: 4
-INITIAL PLANT' CONDITIONS j Generator load:, 350'MWe '
Reactor power: 32% 1 SELECT the statement that describes the expected sequence of events-with-no--
+
oparator actions to increasing pressure' conditions in the main'condenscr. a. The main turbine' will'close, the MSIVs trips will causing a reactor scram, the bypass valves isolate. b. The main turbine the bypass valves trips' close,causing a reactor scram, the MSIVs isolate,- c. The bypass. the main turbine: valves trips,_the isolate. MSIVs isolate causing ~a reactor scram,_ ' d .- 1
.The maina turbine causing trips, the bypass valves close, theLMSIVs isolate reactor scram. ',
i P
' QUESTION: 062 (1.00)~
SELECT the. statement that. describes the limit and-corresponding reason for th3Lnumbers f of manual diesel' generator start attempts from the mainccontrol
. room during a-station blackout.
L
- a. :
Two. attempts during are allowedstarts. the unsuccessful to prevent damaging the journal bearings-
- i. b. One attempt is allowed to prevent depleting the air start E
receivers. c.
~ Two' attempts-are allowed-to prevent depleting-the air start t
receivers. L d .1 Oneiattempt during is allowed tostarts. the unsuccessful prevent damaging the journal bearings L L
-n ,
r REACTOR OPERATOR , a,-- Pcga 4 5' ..i g _ QUESTION: 063 (1.00) 5 q SELECT the statement that describ'es the consequances of failing to complete
.the immediate action of transferring Reactor Feedwater Pump Turbine (RTPT)
- control from automatic to manual for the followi'..g plant conditions. ,
PLANT CONDITIONS DC bus D-1-A is faulted and doenergizcd. Reactor power: 100%. ,
- a. RPV water level continues to increase causing a reactor-scram and damaging the main turbine because the A RFPT and main turbine will' not trip on high level.
b. RPV water level increases until the RFPTs trip on high level, but f the reactor scram will'not occur due to the loss of the DC bus, c. RPV water level increases until the RFPTs trip on high level, the reactor will scram when water-level decreases to '.h>3 level 3'due to loss of feedwater, d. RPV level will increase and stabilize at a higher level when level error overrides the flow error signal on the "A" RFPT. QUESTION: 064 (1.00) SELECT the statement that describes a plant condition that would cause the main turbine generator to' automatically trip.
- a. Vibration is 10 mils for 2 minutes. i b.
Condenser pressure is 5.2 psia and generator load is 500 MWe.
- c. RPV level is +220 inches. I
- d. Exhaust hood temperature is 145F.
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- REACTOR OPERATOR ~ _Pa93;46
.1 QUESTION:=065 -
( l'. 0 0 ) - e SELECT the operator action that will reestablish Control Rod Drive -
. Hydraulic (CRDH) drive differential pressure following a1high drywell pressure scram that cannot be reset. CRDH pump "A" is running. i
- a. Bypass the CRD pump suction filters.
- b. Place the flow control valve in " MANUAL" and open it. l
- c. Fully close the drive water pressure control valve.
d.- Take the high'drywell pressure scram bypass' switches to " BYPASS" and reset the scram and alternate rod insertion (ARI).
' QUESTION: 066 (1.00) ,
SELECT the statement that describes the plant response as RPV' water level
- -increases to Level 8 setpoint. High Pressure Core Spray-(HPCS) and:the-Motor!Feedwater Pump (MFP) were manually started during a loss of feedwater
.flowitransient and recovered level, i
- a. The HPCS pump will_ trip, the.MFP will operate until manual'ly .
secured.- q o b. The HPCS injection valve will close~and'the MFP will trip.
- c. The HPCS pump _and MFP will trip.
'd. The HPCS injection valve will close, the MFP will operate until manually secured.
1 t 1 l l' l l
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l l REACTOR OPERATOR Paga 47 ~ . L
' QUESTION: '067 (1.00)-
SELECTfthe' condition that requires entry into the plant emergency instructions, RPV water level is +182 inches. a. b.- An instrument volume high level signal 4.s present, 25-rods-failed; , to -Jinsert, reactor power is 75/125 scale on IRM' range 4.
- c. Drywell' temperature is 130 F.
- d. Suppression pool; level ~is 19.5 feet. i QUESTION: 068 (1.00) t LGiven the following plant conditions: ,
. 9 PLANT CONDITIONS ;
Containment pressure: '1.4 psig. Containment temperature: 170 F. T
' SELECT the reason containment. spray initiation is prohibited.
- a. The evaporative cooling ef fect :bs not' ef fective. below containment J temperatures of 185.F.
b.. The containment should not be sprayed until, temperature is-greater
.than 185 F to maintain the Residual Heat Removal- (RHR) pumps available for core cooling. '
i
- c. The cooling of the containment could cause the containment design !
negative pressure to be exceeded at containment pressures below
~
l'75 psig.
- d. -The containment should not be sprayed until pressure is greater than 1.75 psig to maintain the Residual Heat Removal (RHR) pumps available for core cooling.
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- REACTOR: OPERATOR'
'Pago 48 !
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6 QUESTION:lO69 (1.00) SELECT a'while the operational' operating:at situ'ation that could cause the following indications 85% power. INDICATIONS i l
-Annunciators alarming APRM'A/E UPSCALE ,
ROD BLOCK APRM UPSCALE I RPV-pressure' increasing, a. Maximum combined flow limit fails high.
- b. Pressure regulator, falls open.
- c. ; Pressure regulatoc fails closed.
- d. Bypass valve-fails open. , <
' QUESTION: 070 (1. 00)'
LGivenz the following plant conditions: )
- PLANT-CONDITIONS '
Reactor power: 4%.
; Main. steam lines are isolated.
s , Suppression pool coo).ing is operating (both loops'). SELECT,'the:
'StCndby. MAXIMUM Liquid Control Suppression Pool temperature-that can exist BEFORE- '
(SLC) MUST be injected. PEI-B13, Figure 1, " Boron Injection Initiation' Temperature" is provided as Attachment 9. a .- 121 F. ,
- b. 120:F.
- c. 110 F
- d. 109 F.
p l: L l
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REACTOR OPERATOR PEgo'49 l QUESTION:~071 (1.00) ; JSELECT the operator action that is required prior to' evacuating the control i rtom if capable'of being performed. t
-a. Place the reactor mode switch in " SHUTDOWN."
;b.
Place all MSIV control switches to "CLOSE."
-t
,c.
-Transfer. station" loads from the auxiliary to startup transformer.
I
- d. Start the division 1 and 2 diesel generators.
4 --QUESTION: 072 (1.00) i SELECT the action that is required to be performed 11mmediately following.a ' total loss of Nuclear Closed = Cooling Water. >
- a. Trip the Recirculation pumps immediately,
- b. Trip the Recirculation pumps within three to five-minutes,
- c. Trip the Recirculation pumps if a high temperature alarm is '
received.
- d. Shift Recirculation pumps from-fast to slow speed.
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I hlbh ,' Paga_50-ii REACTOR OPERATOR: f 0ii > 'l g , T-IQUESTION:LO73-_ (1.00) l-r Given.the following plant conditions: PLANT CONDITIONS t
- Annunciators:=
-SERV AIR COMP TRBL nTBCC HX OUTLET TEMP'HIGH NCC.HX OUTLET TEMP HIGH >
OUTLET TEMP HIGH (stator cooling water) ' RWCU isolation valve, 1G33-F004, isolates Offgas glycol refrigeration compressors trip. , SELECT'the malfunction that caused the conditions to occur. ;
- a. Loss of Turbine Building Closed Cooling Water.
- b. Loss'of Service Water,
- c. Loss of Service and/or Instrument Air.
- d.- Loss of Nuclear Closed Cooling Water.
.e 3 QUESTION: 074: (1.00) l
> . SELECT the sta'tement.that describes a condition requiring a fast reactor
'W .chutdown following a loss of instrument air while operating at 45% power. L a. The 1 SCRAM VLV AIR HEADER PRESSURE LOW annunciator alarms. L b. The ROD' DRIFT annunciator (for rod 22-11) and CRD MECHANISM TEMP 1: HIGH annunciator (for rod 46-39) alarms.
- c. The' outboard main steam line isolation valve (MSIV) B21-F028A
([ ' drifts closed. d
- d. The INST VOL NOT DRAINED annunciator albre.s.
l
h REACTOR OPERATOR
- Pdgo 51-1 i
- QUESTION: 075 (1.00) .;
j
-SELECT the statement that identifies a condition that would NOT be indicative-of a loss of the service /instrumerit air system while operating- l st rated conditions.
-a. Steam seals on the main turbine are lost.
- b. Steam Jet Air Ejector (SJAE) valves close.
- c. Hotwell Pumps trip.
d; Condensate Booster Pumps (CBPs) trip. 1
)
QUESTION: 076 (1.00) SELECT the statement that describes an approved alternate method of decay , heat renoval that can be used if the normal shutdown cooling lineup is lost cnd cannot be reestablished while in operational condition 4. Operats High Pressure Core Spray (HPCS)-to circulate coolant
~
a. between the suppression pool and the reactor vessel via the head
-vent.
- b. Operate Low Pressure Core Spray (LPCS) to circulate t he coolant between the suppression pool and the reactor vessel with LPCS via a: Safety Relief Valve (SRV) .
- c. _ Operate the condensate system to circulate water between the main condenser _and the reactor vessel with the hotwell pumps via the Main Steam Isolation Valves (MSIVs).
- d. Operate Low Pressure Core Spray (LPCS).to maintain level while dumping to radwaste from Reactor Water Cleanup (RWCU).
i
REACTOR OPERATOR Pago 52 QUESTION: 077 (1.00)
'Given=the following . initial-plant conditions:
INITIAL PLANT CONDITIONS Bottom head drain temperature: 180 F. Recirc loop "A" suction temperature: 188.F. RPV level: +235 inches. RPV head vent: Open. Residual Heat Removal (RHR) loop "A": Operating in Shutdown Cooling.- SELECT the statement thttt describes the concern regarding-the "A" RHR pump. tripping.while operating in shutdown cooling mode,
- a. In condition 4, one loop of shutdown cooling must be in service ati all times, therefore this constitutes a Technical Specification violation. .i
-b, The coolant may heatup without adequate monitoring capability, and the plant may. shift from condition 4 to condition 3.
- c. Oxygen may enter the vessel through the open head 1 vent;causingL increased stress cracking corrosion.
- d. The. fuel clad may undergo a zirc-water reaction releasing. hydrogen gas into the containment, i
l l l i l 1 "2
PCg3 33
$ICTOR CPERATOR QUESTION: 079 (1.00)
Giv:n the following plant conditions: PLANT. CONDITIONS Reactor pressure is 855 psig. . Control rod 22-11 is at position 00, its nitrogen accumulatcr has a cracl:ed weld and is isolated for repair.
. SELECT the action required if the operating Control Rod Drive (CRD) pump trips and a'CRD pump CANNOT be restarted. The CRD LEVEL HI/ PRESS LOW Cnnunciator is received for the following rods and the PPO has verified cccumulator pressure at 1350 psig for the alarming HCUs.
Rod Position 18-27 00 38-23 48
- a. Immediately place the reactor modo switch to " SHUTDOWN."
- b. If any other accumulator becomes inoperable for a withdrawn rod, t
immediately place the reactor modo switch in " SHUTDOWN."
- c. If any other accumulator becomes inoperable for a withdrawn rod, start a CRD pump within 20 minutes or place the reactor mode switch in " SHUTDOWN."
- d. Start a CRD pump within 20 minutes or place the reactor mode switch in " SHUTDOWN."
o a
REACTOR CPERATOR P0g3 54 f l QUESTION: 079 (1.00) !
-SELECT the statement that describes the expected plant response following a i rafueling .*mcident where an exposed fuel bundle is dropped and is damaged. l The control room annunciation is listed below. ,
CONTROI. ROOM ANNUNCIATION , CNTMT VENT EXH RAD HI l CNTMT VENT EXH RAD A/D HI HI/INOP l CNTMT VENT EXH RAD B/C HI HI/INOP ;
- a. Vent supply and exhaust dampers remain open, the supply and l exhaust fans trip on the high radiation signal,
- b. Vent supply and exhaust dampers isolate and the supply and exhaust !
fans trip when the dampers isolate,
- c. Vent exhaust dampers isolate and the exhaust fans trip when the dampers isolate, the supply fans continue to operate.
- d. Vent supply dampers isolate und the supply fans trip on the high radiation signal, the exhaust fans continue to operate.
P QUESTION; 080 (1.00) [ SELECT the statement that describes an approved method for emergency d3 pressurization in accordance with PEI-B13, " Reactor Pressure vessel Centrol." t
- a. Open any one of the Safety Relief' Valves (SRVs).
- b. Open any two Automatic Depressurization System (ADS) valves.
- c. Depressurize the reactor by opening the head vent if only one Safety Relief Valve (SRV) can be opened.
- 4. Depressurize the reactor using Residual Heat Removal (RHR) ster, ,3 condensing node if no Safety Relief Valve (SRV) can be opened.
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, REACTOR CPERATOR Paga 35 ,
p
. QUESTION: 081 (1.00) ,
3 SELECT the statement that describes the Redundant Reactivity Control System ' rocponse that will occur 25 seconds after pressure exceeds 1083 psig following a failure to scram on a main turbine trip at 45% power.
- a. Feedwater flow is reduced to zero.unless the feedvater controllers are in manual.
- b. Alternate rod insertion will actuate and can be reset 30 seconds i later.
- c. The reactor recirculation pumps will trip.
- d. Standby Liquid Control will automatically initiate.
QUESTION: 082 (1.J0) SELECT the statement that describes an operator action required for the , following plant conditions. PLANT CONDITIONS Reactor power: 75%. Suppression pool temperature: 131 F. Suppression pool level: 18 feet 2 inches. SRV 1B21-F047B: Failed open.
- a. If the SRV cannot be closed within five minutos place the reactor mode switch in "SHUTDOW:!."
- b. Immediately place the reactor mode switch in 'dHUTDOWN."
- c. Reduce suppression pool '.evel to less than 18 feet,
- d. 'If suppression pool te.nperature exceeds 120 F, arm and depress the manual scram pushbuttoca.
l l l t [
Paga 56 REACTOR OPERATOR 4 I QUESTION: 083 (1.00) Giv;n the following plant conditions: PLANT CONDITIONS Suppression pool temperature: 192 F Suppression pool level: 16 feet. Containment pressure: 2.1 psig. SELECT the limit on operating the Reactor Core Isolation Cooling (RCIC) Cyctem. PEI-E12, Figure 1, "RCIC Turbine Speed Limit" is provided as-AttCchment 11.
- a. The RCIC turbine should not exceed 2000 rpm.
- b. The RCIC turbine should be operated to limit pump flow to 700 gpm.
- c. The RCIC turbine should not be operated.
- d. The RCIC-turbine'should be operated at reduced speed to limit. pump flow to 350 gpm.
1 i l l
{ REACTOR OPERATOR P093 57 . l QUESTION: 084 (1. 0 0) . GivGn the following plant conditions: - PLANT CONDITIONS !
.Drywell temperature: 325 F.
Containment temperature: 220 F. j i RPV pressure: 100 psig. SELECT the statement that describes an RPV water level instrument that may b3 used for trending the reactor's water level. PEI-D23-3, Figures 1, 2b, 2c, 2d, and 2e, are provided as Attachment 12.
- a. Wide range level indicates +9 inches. l
- b. Upset range level indicates +190 inches.
- c. Shutdown range indicates +200 inches.
- d. Harrow range indicates +170 inches. i QUESTION: 085 (1.00)
SELECT the statement that describes the reason emergency depressurization l 10 raquired if containment temperature cannot be maintained less than l l 185 F. V
- a. Maintain equipment operability by reducing further energy addition -
from the reactor to the containment. ; p b. Operability of the safety relief valves (SRV) is in jeopardy and. - it-is appropriate to emergency depressurize before they fail. ,
- c. The suppression pool may no longer be able to absorb the energy i from a loss of coolant accident.
- d. Containment liner failure may be imminent and it is necessary to place the reactor in a low energy state.
L ' L 4 1
Pcg3 S8 REACTOR CPERATOR i l QUESTION: 086 (1.00)
' SELECT the statement that describes the reason reactor vessel water level instruments may not be used for level indication if the drywell temperature n ar the reference legs is greater than the RPV Saturation Temperature.
- a. The differential pressure transmitters are not environmentally qualified to operate at soturated temperature conditions. l
- b. Drywell pressure is the same as reactor pressure providing a zero differential pressure (upscale level indication).
- c. Actual RPV water level may be lower than indicated water level.
- d. The density of the water in the reference leg is too low to provide a usable differential pressure to measure level.
i REACTOR CPERA7tR P0go 59 ! i QUESTION: 087 (1.00) , Given the following plant conditions: PLANT CONDITIONS i Reactor pressurch . 950 psig. Suppression pool level: 26.5 feet. , Suppression pool temperature: 150 F. ; SELECT the action required and its corresponding reason. PEI-G42, Figures , 1.cnd 2, " Heat Capacity Level Limit" and " Suppression Pool Load Limit" are I provided as Attachment 13.
- a. Emergency depressurize the reactor to prevent excessive ruppression pool dynamic loading due to SRV actuations. l
- b. Evergency depressurize the reactor to ensure the suppression pool :
can absorb the energy released by a LOCA.
- c. Reduce suppression pool level to less than 23 ft. to prevent containment design temperature from being exceeded if an ADS blowdown were to ocrur, j
- d. Reduce reactor pressure to less than 800 psig to prevent exceeding !
the containment design temperature in the event of a Design Basis Loss of Coolant Accident.
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REACTOR OPERATOR PCgo 60 i l QUESTION: 088 (1.00) SELECT the statement that describes the reason for requiring suppression poal level to be greater than 5.25 feet prior to commencing emergency 1 d: pressurization. ,
- s. The Safety Relief Valves (SRVs) tailpipe would be uncovered and ,
discharge directly to containment, the containment design pressure could be exceeded.
- b. The submergence of the Safety Relief Valves (SRVs) is insufficient for complete exhaust steam condensation and the containment design pressure could be exceeded.
- c. The reduced mass of water that is cleared from the tailpipe'when the Safety Relief Valves (SRVs) is opened would place excessive velocity loading on the tailpipe.
- d. To provide sufficient back pressure on the Safety Relief Valves (SRVs) to ensure the SRVs will reclose when the control switches are placed to "CLOSE."
QUESTION: 089 (1.00) , SELECT the alternate injection system lineup addressed by PEI-B13, " Reactor
- Prcssure Vessel Control" that could be used in the event reactor water lovel CANNOT be maintained greater than zero inches. [
- a. Standby Liquid Control (SLC) transfer pump injecting via the reactor water cleanup (RWCU) return line.
- b. Emergency Service Water (ESW) injecting via the Residual Heat Removal (RHR) shutdown cooling to feedwater line.
- c. Emergency Service Water (ESW) injecting via the High Pressure Core
' Spray (HPCS) injection line.
- d. The fire water system injecting via the Low Pressure Core Spray (LPCS) injection line.
i Pogo 61 REACT ^R OPERATOR i t QUESTION: 090 (1.00) , SELECT the statement that describes a situation where adequate core cooling 10 Cssured. PEI-B13, Figure 4,
" Minimum Alternate Flooding Pressure" ic !
provided as Attachment 14.
- a. Reactor power is 7%, two Safety Relief Valves (SRVs) are open, RPV ,
water level is -11 inches, reactor pressure is 450 psig, Reactor Core Isolhtion Cooling (RCIC) is injecting. ,
- b. One safety relief valve (SRV) is open, RPV pressure is 100 psig, RPV level is unknown, Residual Heat Removal (RHR) loop "A" iu injecting.
- c. Two Safety Relief Valves (SRVs) are open, RPV water level is
-50 inches, RPV pressure is 40 pJig, no injection systems are -
operating.
- d. One safety relief valve (SRV) is open, RPV pressure is 800 psig, RPV water level is -135 inches, no injection is available. ,
(1.00)
-QUESTION: 091 SELECT the reason reactor power decreases as RPV water le"el is lowered >
during and ATWS.
- a. The single phase natural circulation flowpath through the steam separators is broken causing the remaining water inside the shroud ,
to boil and totally void the core region.
- b. The driving head from the downcomer water level is reduced, minimizing core flow, increasing the voids in the core region,
- c. Carryunder increases, increasing the preheating of the coolant, reducing the core inlet subcooling of the coolant.
-d. The total mass of coolant in the reactor vessel is reduced, causing it to rapidly void the entire core region.
l
. REACTOR OPERATOR Pago 62 i l 1 1 >
QUESTION: 092 (1.00) i While operating at 100% power, the following conditions occurt PLANT CONDITIONS Annunciators ANNULUS A DIFF PRESS LOW ANNULUS B DIFF PRESS LOW , Annulus differential pressures Zero inches of water gage SELECT the potential adverse consequence that could occur if this situation ' is not corrected.
- a. Access to containment could be limited due to the differential !
pressure across the access doors. t
- b. 'The annulus region could be overpressurized in the event of a Loss of Coolant Accident (LOCA)
- c. A ground level, unmonitored release could occur in the event of a i design basis Loss of Coolant Accident (LOCA).
- d. Excessive load would be placed on the containment vent and purge systen due to increased inloakage to the containment. -
QUESTION: 093 (1.00) , SELECT the statement that describes an action required in order to reset a rOLctor scram during an ATWS.
- a. Place the LO POWER SET PT DIV 1 and 2 BYPASS keylock switches to
" BYPASS."
- b. Place the INST VOL LEVEL HI SCRAM BYPASS keylock switches to
" BYPASS."
- c. Arm and depress the RRCS MANUAL ARI pushbuttons.
- d. Place the reactor mode switch to "RUN."
T*
REACTOR CPERATOR PCg3 63 QUESTION: 094 (1.00) Th0 reactor is in operational condition 2 during a startup. Power is in thJ heating range and reactor water level +200 inches. When turbine shell warming is commenced, RPS initiates a reactor scram. SELECT the cause of the reactor scram.
- a. Turbine first stage pressure indicated greater than 40% power.
- b. Reactor water level decreased to level 3 when steam demand increased.
- c. The MSIVs closed due to low reactor pressure when stean demand increased.
- d. Reactor power increased due to void collapse.
(********** END OF EXAMINATION **********) , i
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3036-E21(NS4)-00. Obj. 0 .
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-Perry System Description Manual Chapter B21/N11 OT-3036-921/N11-01. Obj . C P(3.9/3.93
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Perry System Description Manual Chapter N21. Chapter til CRDH 1' OT-3036-C11(CRDH)-02. Obj. B.- -
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q - Perry System Descripii ^ Manual Chapter B33 OT-3036-B33-01, Obj.
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r 6 OT-3036-R14/R15, Obj. B, G Perry System Description Manual Chapter R14/R15 103.1/3.43 262002K401 ..(KA's) t IhNSWER: 053< (f.00) .j y
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.0T-3036-R43/48-01, Obf.et.ves C, H.
Facility-examination +;.,n question M3092 Perry System _ Description Manual, Chap. R43 .
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' REFERENCES-OT-3036-N64-01, Obj. E, K. OT-3036-D17A-01. Obj. D Perry System Description Manual Chapter N64 (3.0/3.2)- I
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REFERENCES ; ONI-B33-2. Loss of One or Both Recirculation Pumps Facility examination bank question H3072
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ONI-N62, " Loss of Main Condenser Vacuum." !
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REFERENCE:
l. iANSWER:- 066 (1.00)' i
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- PerryTSystem Description Manual: Chapter'E22A. B21 NBPI. .
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REFERENCE:
PEI-B13, " Reactor Pressure Vessel Control."
'PEI-D17, " Radioactivity. Release Control." I I
PEI-042, " Suppression Pool Leve) Control.' [ 4. 2 / 4 '. 5 3 . s
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'i . ANSWER 9 . 068 (1.00) c.
REFERENCE:
.t' OT-3034-D23-1. Obj. C page 8. >
PEI-D23-1, " Containment Temperature Control" step 3.1.6.
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REFERENCE:
ONI-C05-1, " Pressure Regulator Failure-Closed." 4 [3.3/3.41 i 295014A106. . . ( KA 's ). I
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- 0N!-B13. " Reactor Pressure Vessel Control." !
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ONI-C61
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REFERENCE ' q ONI-P41, " Loss of Service Water." ON!-F43, " Loss of Nuclear Closed Cooling Water." ; ON!-P44, " Loss of Turbine Building Closed Cooling Water." - ,
-ON1-PS2, " Loss of Service and/or Instrueent Air." :i (3.0/4.13- ,
2950100011 ..(KAfs) ;
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, :I ONI-P52, " Loss o Service and/or ' Ins trument Air."
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ONI-P52. " Loss of Service and/or Instrument Air." [3.9/4.1]' 1: L 2950190011, ..(KA's)
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- %NSWER: 076' (1.00) '
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; .C3.7/3.73 293021A104 ..(kA's) = ANSWER:. : 07 7 -- (1.00). ,
b.. REFERENCE ,
- 0NI-E12-2, " Loss of Shutdoan-Cooling." ,
Perry System. Description Manual Chapter E12. . _' C 3. 9 / 3. 9 3
;295021K102 ..(KA's) i J
I ANSWER: '070 (1.00)L i D i I' o I i 7 8.! ; 1
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REFERENCE:
PEl-B13.;," Reactor Pressure Vessel Control." il PEI-D23-2,-"Drywell and Containment Pressure Control." .;
-(3.0/3.8) 1295024A203 ..(KA's)
ANSWER:- 001 .(1.00)
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l Perry System Description Manual Chapter C22. , [3.3/3.7) ,l _' ~295025A100' . .~. ( K A 's ) a
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. .ONI-B21-1 "SRV Inadvertent Opening / Stuck Open." ,
'!: Perry Technical-Specifications section 3.6.3.1. l (4.1/4.2) t 295026A201 ..(KA's) ,
t ANSWER: 083 (1.00)' s c. REFERENCES-
-i PerryfSystem' Description Manual Chapter E51.
/' .0T-3034-02-E12-00,.0bj. C.
PEI-E12, " Suppression Pool. Temperature. Control." [3.0/4.53'
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1 004: (1.00) i
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PEI-D23-2, "Drywell Temperature Control." [3.0/3.2) ,
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[REFEREtJCE ' , OT-3034-02-D23-1, Obj. B. ; [3.7/3.8) 295027K301 ..(KA's)- i
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PEI-042, " Suppression' Pool Level Control." !
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ANSWER: 000 (1.00) '
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PEI-B13 " Reactor Pressure Vessel Control." [3.6/3.9) 295030G007' ..(KA's) ANSWER: 009 -(1.00) { i
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REFERENCE:
-PEI-B13. "Rea-tor Pressure Vessel Control."
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01-3034 A1 T S-00. Obj . C. (3.7/4.13 I "295031K103I ..(KA's). _
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t PEI-B13, " Reactor Pressure Vessel Control." ,; [4.2/4.33-
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~0T-3036-C71-02 Obj. E.- , ;
(3.b/3.0) , 29 BOO 6A206 ..(KA's) 6 I t
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.L U. ' S. NUCLEAR REGULATORY COMMISSION
~
. SENIOR REACTOR 10PERATOR-LICENSE EXAMINATION REGION 3-FACILITY:' Perry li& 2-n REACTOR TYPE: BWR-GE6 DATE ADMINISTERED:, 90/08/06 CANDIDATE: j
- t
') e 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 pape;stwill be picked up four and one half (4 1/2) hours after the. examination starts. .
.k
-NUMBER TOTAL- CANDIDATE'S . CANDIDATE'S
. QUESTIONS- POINTS POINTS. OVERALL oGRADE-(%)
95 98.00
,i
' All work'done on this examination is my own,- I have-neither given
' nor-received aid.
Candidate's signature i. U mSTER COM :
't -
4 .- ,
w j NRC RULES AND: GUIDELINES FOR LICENSE EXAMINATIONS l
- During theladministration of this examination the following rules apply:
- 14 Cheating on the: examination means an automatic denial'of your application.
cnd;could result.in more severe penalties. 8 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 l done after you complete the examination. ,
i
- 3.:Restroom trips are to 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, j
'4 . Use black ink or dark pencil only to facilitate legible reproductions.-
5 3 Print your name in the blank provided in the upper right-hand corner of the examination. cover sheet. t
- 6. Fill in the date on the cover sheet of cne examination (if necessary).
- 7. You may write.your answers on the examination question page or on a- I ceparate sheet of paper. USE ONLY THE PAPER PROVIDED AND DO NOT WRITE ON-
'THE BACK SIDE OF 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 separatetsheet of the paper provided and insert it directly after the specific question. DO NOT WRITE ON.THE BACK SIDE OF THE EXAMINATION QUESTION PAGE. E ;9. Print your name in the. upper right-hand corner of the first page of answer
. cheets whether you use the examinc. tion question pages or separate sheets- ,
.of paper. Initial each of the following answer pages. }
p 10. Before you turn in your examination, consecutively number each answer cheet, including any additional pages inserted when writing!your answers I
. on the examination questien page.
T11'. If you'are using separate sheets, number each answer and skip at least 3 L lines'between answers to allow space for grading. (~ 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 .: crror=resulting:in an incorrect answer. Write it out. ' i (- - - - --" -- _.- - _ _ -
m
- j" >
14( Thc. point-volu3 for occh qusction 10 indic tcd in paronthecos ofter tha question.. The amount-of blank space on an examination question page is
' NOT an indication of the depth of answer required. ,
akb.Showall: calculations, methods,orassumptionsused-toobtainan' answer. ; L16.L P2rtial: credit may be given. Therefore, ANSWER ALL PARTS OF THE QUESTION- , r-AND i DO NOT LEAVE - ANY ANSWER - BLANK. . NOTE:~ partial credit will NOT be 4
. given on multiple choice questions.
;17 % Proportional grading will be applied. Any' additional wrong'information i thatois provided may count against you. For' example, if a question is ~
- worth one. point and asks.for four responses, each of which is worth 0.25 points, and you give five responses, each of your responses will be worth 0.20 points. If one of your five responses is incorrect, 0.20 will be '
daducted and your total credit for that question will be 0.80'instead of
' 1.00<even.though you got the four correct answers.
718.JIfithe intent of a question is unclear, ask questions of the examiner
; - only.
L19.-When turning in your examination, assemble the completed examination In with addition, 4
,c examination questions, examination aids and answer sheets. ;
turn in all scrap paper.
' 20.1To pass the examination, you must achieve an overall. grade of 80% or greater.
521. There is a time limit of (4 1/2) 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-otion 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. i l< (h y 1.
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. SENIOR REACTOR-OPERATOR =Pags 4 t, ~!
L QUESTION: 001 (1.00)
= ?
SELECT.the statement that describes the proper' method of verifying the position of a valve after initial positioning of'the valve has been , c mpleted fol. lowing maintenance on the valve.' The valve checklist. required ptaition is "LX." L a '. - Verify-it closed by position indication and resistance to turning' the handwheel in the closed direction. < Verify it' closed by turning the handwheel 1/8 to 1/4 turn open and_ b. reclosing the valve, attach the locking device.
.i ic. Verify it closed by turning'the handwheel 1/8 to 1/4 turn open and reclosing the valve.
- d. Verify the valve closed by position indication and resistance to turning the handwheel in'the closed direction, attach the lock 2ng- -
. device. ;
I
)
7 h 4 3 1 1 i I 4
+-
NENIOR REACTOR OPERATOR; Paga 5
. QUESTION: 002 (2.00)-
MATCH each of-the situations in Column A with the type of tag.that would be , ucsd in Column B. The MINIMUM' tagging required for each situation should-bo identified. Each item in Column B may be used once, more than once, or not at all. ONLY a SINGLE answer is used for each item in Column A. COLUMN A COLUMN B (SITUATIONS) (TYPE OF TAG) a..A valve that could cause 1. Jurisdictional' Tag - personal injury if operated. 2. White, Out of Service
- b. A valve undergoing minor 3. Red Tag maintenance lasting 'ess than one shift. 4. . Yellow Tag c.-A red-tagged circuit 5. MFI Tag breaker that could provide power to a grounded bus. 6. LLJED Tag
- d. A blocked relay in a logic 7. ILRT/LLRT Tag train.
U p Paga: 6. l SENIOR 1 REACTOR OPERATOR L LQUESTION:-003 (1.00)
' SELECT lthe statement that describes a condition-for accomplishing an >
1
-EXPEDITED tag clearance.-
- a. _The " person-in-charge" individuals of the original work order and ,
lany riders must approve clearing the. tags. +
' b.. .The managers.or supervisors:of the " person-in-charge" individuals- ;
must' recall or stop all work associated with-the tags being + cleared and concur with the clearance, n c '. The Unit Supervisor may authorize removal of the tags, but must
/ inform the " person-in-charge" individuals as soon as possible. ,
- d. Attach messages to the badges of the " person-in-charge" individuals: informing them of the cleared tags.
y QUESTIO'N : 004 ..(1.00)
- SELECT the statement that describes the quarterly whole body' (WB). exposure ,
1imit'forlan individual with the. exposure' history identified-below as per. PNPP administrati've guidelines.
;EXPOSURR HISTORY Age:
.25 years.- 1 Lifetime exposure: unknown.
lNRC--Form 4: not available, exposure history is. undocumented.
- C.iM p Id bp(M.- C vr&. 1
!a. l300. mrem. .
4 Lb.- 1000 mrem.
- c .- E1250 mrem.
d.' -3000. mrem. , s 4 I .
'n t 4 . _ _ - _ _ _ _ . _ _ _ _ _ . _ . _ _ _ _ _ _ , , . . _ _ f
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qG J!, t JSENIOR REACTOR OPERATOR- Pcgc_ 7-
'W l 3
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; QUESTION:'005 (1.00)
(SELECT the statement that describes the remaining Federal'whole body- (WB)
,, ..Gxposure for the current calendar quarter for an individual with the
, ? exposure history. identified below. '
L EXPOSURE HISTORY
. Age: 22 years. $
Lifetime exposure: 21 rem. Current quarter: 250 mrem. NRC Form 4: On file, lifetime and current quarter exposure history _ is documented.
- a. 250 mrem.
t
'b. 750 mrem.
.c. 1000' mrem.
d.. 2750 mrem. r t
; QUESTION:1006' (1.00)
SELECT the statement that describes a circumstance where a radiation work p3rmit - (RWP) -is' required. b
.a. Areairadiation levels are 75 mrem / hour.
- b. Airborne ~ activity levels are 0.2'MPC due.to noble gas.
c. Area ~ neutron radiation levels are'O.25 mrem / hour.
, d. -Anjob_where a worker will receive 80 mrem in one shift.
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,. SENIOR REhCTOR OPERATOR Pcgg" 8
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i t EQUESTION: 007 (1.00)- SELECT the-statement that describes a responsibility,of any radiation
.hc worker at the Perry plant.
- r
- a. Surveying high radiation areas entered while on a radiation' work ,
permit. ' Complying with all of the requirements identified on the radiation-l b. work permit for a controlled area to be entered. t Evacuating a controlled area if their self-reading dosimeter
~
c. reaches 50% of its maximum reading, , i
- d. Refusing to enter controlled areas with general' radiation-levels u -in excess of 150 mrem / hour. ,
w s t v i;'! QUESTION:008 (1.00) I SELECT the statement that describes the MINIMUM personnel protective 1 equipment required in addition to eye protection-when racking out the I K .L1205, Recirculation Pump "B" fast speed breaker. s i f .a. , Leather work gloves. ? } ib. 10'Kv rubber gloves and leather gloves.
- c. 20fKv rubber gloves and leather gloves. ;
g d. 20 KV rubber gloves, leather gloves, and 20 Kv rubber sleeves. c. i &.,u : lQ a !. k r t 1
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e 4 t .~ [ SENIORiREACTOR OPERATOR Paga 9 1 k QUESTION:1009 _( 1. 00) . 1: b SELECT the statement that describes a situation that requ' ires the. Unit
- Supervisor'sf approvalt for entryfi nto c confined space.
La. Morel than three individuals at one time-will be.in the space.
;b.
Cutting torch operations will be occurring in the space, q c. An' oxygen deficient atmosphere is present in.the space,
- d. An explosive atmosphere is present in the' space..
= QUESTION: 010 (1.00) !
i SELECT Lfir@' the plant brigade team. staff member who is . excluded from being a . member cNE .the-a.- ' Supervising Operator (not at the controls). ]-
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b.- Non-licensed. operator, i
- c. Rad-waste supervisor.
- d. ' Shift Supervisor.
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a a SENIOR REACTOR: OPERATOR _ -Pagt 10 k i
" QUESTION: 011. (1.00) 3 SELECT the: statement that is a requirement for the conditional approval of.
La "non-intent" temporary change. s
.a.- It is not' desired for the procedure change to become permanent.
- b. The technical and-maintenance section managers both approve the ,
temporary change. ;
- c. The Unit Supervisor approves the change.
- d. Final disposition is not completed for 16 days frcm the effective date.
(1.00) QUESTION: 012
' SELECT the statement that describes a situation that does NOT' meet the
.rsquired shift manning in the control room when the plant is operating at 85% power under, steady-state' conditions. Each statement lists where each person is located. -
- a. Shift Supervisor: Pump house f Unit Supervisor: Back panels j Supervising Operator at'the controls: Horseshoe area Second: Licensed: Operator: Back panels,
- b. Shift. Supervisor: Operations manager's office Unit Supervisor: STA office
; Supervising Operator at the controls: Back panels Second Licensed Operator: Containment ~ building.
- c. shift Supervisor: Administrative. building Unit Supervisor: STA office Supervising Operator at the controls: Horseshoe area
!Second Licensed Operator: Containment building.
- d. Shift Supervisor: Technical Support Center Unit-Supervisor: Unit Supervisor desk; Supervising Operator at the controls: Surveillance area.
Second Licensed Operator: Containment building.
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, sa F hi~~ ..?' SENIOR; REACTOR OPERATOR: ;- Paga_11-
.o 1 l l QUESTION: 013 (1.00)
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.Given thetfollowing surveillance data:
-e
; ) , ~~ .
}' SURVEILLANCE DATA-i
-Recirculation pumps are in' fast speed. i
;1B33-R613: .m 45 M1bm/hr.
JetLpump data.(Mlbm/hr): i 1B33-R611A PS : - ~1.6- t 1B33-R611B P10: 1.7 1B3 3-R611'C',0 JP15 : 1.7? .. 1833-R611D,-JP20: 1.6-APRM flow readings (%): ,, o Div. 1' A: 39 E: 39 i Div. 2 :' B: 36 'F: 36 . Div.:3: 'C: 40 G: 41 ' Div.:4:- ;D: 38 H:-38 , n ' SELECT the statement that describes the results of the performance of. '
- SVI-C51-T0026, "APRM Finw-Biased Power-Flow Verification.," Copies.of c A003,1 1
'"Pdrcent Drive: Flow.ys Total Core ~ Flow" and-SVI-C51-T0026 are provided as ,
'" Attachment 1.--
x-
.a._ Both;div.;1 and-both div. 3 APRMs are-inoperable.
>b. The "G" APRM is inoperable.
ql c. Both div.-3 APRMs are inoperable. n ~ '
'd. Surveillance is satisfactory. _
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' SENIOR REACTOR OPERATOR' Paga 12 I'
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; QUESTION: 014 (1.00) '
i SELECT therstatement that identifies'the most RESTRICTIVE action-required
.btsed on the primary: coolant sample results listed below. The reactor is- l
; operating at 85%' power and1the previous sample results were all normal. ,
The applicable sections of Technical Specifications ar.2 Ch0mistry Control" are provided as Attachment'17 and 2. P,'P-1102, " Plant j PRIMARY COOLANT SAMPLE RESULTS ~ Conductivity: Chloride: 1.1 micro-mhos/cm 0.25-ppm-
-Dose equivalent iodine:
Total activity: 1.9 micro-curies / gram - 85/E-bar micro-curies / gram
- a. Reduce conductivity to less.than 1."
micro-mhos/cm within 24 houcs or conduct an orderlyfshutdown'to. cold-shutdown within 16 hours. ' b. Commence an orderly shutdown to cold shutdown within 16 hours. . c. Be in24 next hothours. shutdown within 12 hours.and cold shutdown within the t
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d. Be in hot shutdown with the MSIVs closed within the next 12 hours.
-QUESTION: 015 (1.00)
SELECT the statement that describes a responsibility:of the Emergency Coordinator that may be delegated _during a Site Area Emergency.
.a. Downgrading the emergency to an alert-level, b.
Directing the activation of - the technical support center (TSC) . c. Deciding what officials. protective actions to recommend to Ashtabula county
.d.. Deciding to notify the State of Ohio officials.
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IT ! L SENICR REACTOR OPERATOR-P;ga 13 f
'bUESTION:.016
; (1.00) '
SELECT (LOOP) withthe emergency the followingclassification required for a loss of off-site power plant conditions. EPI-A1, " Emergency Action Levels" Attachments 1, 2, and 3 are provided as Attachment 3. , PLANT CONDIU,0NS TOLLOWING THE LOOP Reactor mode switch position: RETUEL Core offload is in progress, 50% of bundles in-core. Division II and Division III diesel generators are out of service for maintenance. Division I diesel generator fuel rack trips on its initiation and 25 " minutes elapse before it can be manually started and tied to its hus. Division LOOP. II diesel generator is tied to its bus 95 minutes after the
- a. Notification of Unusual Event. i
- b. Alert.
- c. Site Area Emergency.
- d. General Emergency.
QUESTION: 017 (1.00) P
" SELECT-the statemont that describes the operation of the Redundant R32ctivity Control System's (RRCS) Alternate Rod Insertion (ARI) valves wh n ARI actuates. ,
- a. Energize from the 125 Vdc bus EDIA (1R42-S012).
- b. Deenergized from the 125 vde bus E01A (1R42-S012).
c. Energize from the Reac*or Protection System (RPS) bus B.
- d. Deenergize from the-Reactor Protection Systein (RPS) bus B. ,
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[ } i . SENICR REACTOR OPERATOR Paga 13 { 4 QUESTION: 016 (1.00) , SELECT (LOOP) withthe emergency the followingclassification required for a loss of off-site power plant conditions. Levels" Attachments 1, 2, and 3 are provided as EPI-A1, " Emergency Action Attachment 3. . PLANT CONDITIONS TOLLOWING THE LOOP
- Reactor mode switch positlent RETUEL Core offload is in progress, 50% of bundles in-core.
' Division II and Division III diesel generators are out of service for maintenance. -
Division I diesel generator fuel rack trips on its initiation and 25 i minutes. elapse before it can be manually started and tied to its bus. ! i Division II diesel generator is tied to its bus 95 minutes after.the LOOP. , l
- a. Notification of Unusual Event. t
- b. Alert.
- c. Site Area Emergency. i
- d. General Emergency.
QUESTION: 017 (1.00) 1 SELECT the statement that describes the operation of the Redundant ROCetivity Control System's (RRCS) Alternate Rod Insertion (ARI) valves whOn ARI actuates.
- a. Energize from the 125 vde bus EDIA (1R42-S012).
1
- b. Deanergized from the 125 vde bus EDIA (1R42-S012).
'c.
Energize from the Reactor Protection System (RPS) bus B.
- d. Deenergize from the Reactor Protection Sys.em (RPd) bus B.
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r SENIOR: REACTOR. OPERATOR p 93 14 L l' r : u :n e m i 1 QUESTION: 018' (1.00) g !, 'rod SELECT the statement that describes the' indications that verify a control'
- 48. is coupled to its drive mechanism when it is notched out from position !
t ' a.
" FULL OUT" LED remains lit, " ROD OVERTRAVEL" annunciator does not alarm. ^I '
?
- b. "TULL OUT" LED goes out, " ROD OVERTR/./EL" annunciator alarms and-immediately clears. ,
t
- c. h Rod position indication goes blank then indicates "48" again, " ROD' '!
OVERTRAVEL" annunciator alarms and immediately clears. j I' d. Rod position indication goes blank, " ROD OVERTRAVEL" annunciator alarms. 3 i i L!
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r-l (' .. SENIOR REACTOR OPERATOR P093 15 i l QUESTION: 019 (1.00)
-Giv n the following plant conditions:
PLANT CONDITIONS Reactor power 45% Generator loadt 460 MWe control rods.in group 9 are being withdrawn by notch withdrawal from notch 12 to notch 24 (the desired rod pattern line). ELECT the statement that describes a result of the main turbine first ttge shell pressure input to the Rod Control and Information System (RC&IS) failing to its maximum pressure value, a.- Any additional rod withdrawal is limited to four notches of continuous travel for any selected rod in the allowed sequence,
- b. Any further_ rod withdrawal is limited to two notches of continuous travel for any selected rod in the allowed sequence.
- c. Any further rod withdrawal is NOT limited by notch withdrawal constraints in the allowed sequence.
- d. Any further rod withdrawal of any rod is biccked.
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SENIOR REACTOR OPERATOR . . . Pago 16 QUESTION: 020 (1.00) i A thereactor il Bypassstartup Valveis(BPV) in progress opens.and control rods are reing withdrawn when SELECT the action required given the following plant conCitions. ' PLANT CONDITIONS l Generator output: 250 Mwe. Reactor power: 26% Load limiter setpoint: 250 MWe.'
- a. Close the bypass valve by inserting control rods.
b. Close the bypass valve by increasing the pressure setpoint. l c. Close the bypass valve by increasing the load limiter setpoint. ! d. Close speed. the bypass valve by shifting recirculation pumps t.o low QUESTION: 021 (1.00) i SELECT the statement that describes the effect on the Recirculation system for an actuation turbine trip. of the end of core life (EOC) interlock following a main a. At less than 40% rated power, the low frequency motor generator ' (LFMG) and the CB3 and CB4 breakers will trip for each pump, b. At greater than 40% rated power, the CBS breaker will trip, the low frequency motor generator (LFMG) will start, and the CB2 breakers will close for each pump. l
- c. At greater than 40% rated power, the CB3 and CB4 breakers will trip, the low frequency motor generator (LFMG) will start, and the CB2 breaker will close for each pump.
d. l At less than 40% rated power, the CBS breaker will trip, the low frequency motor generator (LFMG)' will start, and the CB2 breaker will close for each pump.
F l SENI;R REACTOR OPERATOR Paga 17 QUESTION: 022 (1.00) SELECT the statement that describes a situation when the Low Pressure Coolant Injection (LPCI) From RHR A Shut-off Valve (1E12-F042A) will open.
- a. A LPCI signal is present and the downstream loop pressure is GREATER THAN 530 psig. .
- b. A LPCI signal is present and the RHR A injection valve i (1E12-F027A) is open. !
- c. The control switch is taken to open and the blue light is l illuminated indicating downstream loop pressure is LESS THAN '
530 psig.
- d. The control switch is taken to open and the RHR A injection valve ,
(1E12-F027A) is open. ; QUESTION: 023 (1.00) SELECT the statement that describes the potential effect on the "B" and "C" ROcidual Heat Removal (RHR) subsystems as indicated by the receipt of the 4 "RHR B OUT_OF SERVICE" and "RHR C OUT OF SERVICE" annunciators due to the waterleg motor pump (IE12-C003) supply breaker tripping,
- a. The pumps are in danger of electrical overload in the event of a LOCA initiation and must be declared inoperable.
- b. The pumps may not inject to the reactor vessel within their '
analyzed injection time in the event of a LOCA.
- c. The pumps will not start in the event of a LOCA since their supply breakers are interlocked open.
- d. The pumps may not achieve design flow due to excessive pump cavitation.
I SENICR REACTOR OPERATOR Pcga ic QUESTION: 024 (2.00) MATCH the reactor water cleanup (RWCU) system response in Column B to each cf the operating conditions in Column A. The RWCU system is initially in ito normal full-pewer configuration. The responses in Column B may be used once, more than ofice, or not at all. ONLY a SINGLE answer is used for each itcm in Column A. COLUMP A COLUMN B (OPERAfING CONDITIONS) (RWCU SYSTEM RESPONSE)
- a. RPV water level is 1. Only the pump suction inboard
+104 inches, isolation valve (1G33-F001) isolates.
- b. RWCU differential flow is 60 gpm. 2. Only the pump suction outboard
- c. The keylock switch for the isolation valve (1G33-F004) isolates.
"A" Standby Liquid control pump is taken to "ON". 3. Both the pump suction inboard and outboard isolation valves
- d. RWCU non-regenerative heat (1G33-F001 and -F004) isolate.
exchanger (NRHX) outlet temperature is 132 F. 4. No isolation action occurs.
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SENIOR REACTOR CPERATOR p ga 19 ,
" I i
g, QUESTION: 025 (1.00) 1 Civen.the following initial plant conditions: INITIAL PLANT CONDITIONS i
, RHR loop "B" is -in Shutdown Cooling (SDC) mode.
Coolant temperature is 300 F. RPV pressure is 65 psig. i SELECT the statement that describes the effect on the Shutdown Cooling ; (SDC) Suction Isolation Inboard and Outboard Valves (1E12-F009 and 1E12-F008) if the 4.16 KV bus EH12 trips. :
- a. 1E12-F008 and 1E12-F009 vill not isolate at 135 psig due to a loss.
of power. ,
- b. 1E12-F008'and 1E12-F009 will both isolate when pressure reaches 135 psig.
- c. 1E12-F009 isolates when pressure reaches 135 psig. 1E12-F008 will not isolate due to a loss of power. ,
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- d. 1E12-F008 isolates when pressure reaches 135 psig, 1E12-F009 will not. isolate due to a loss of power.
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SENIOR REACTOR OPERATOR Pag) 20 QUESTION: 026 '(1. 0 0) , .Civen the following plant conditions: ; PLANT CONDITIONS RPV water level:' +16 inches for 105 seconds. Drywell pressure 2.0 psig l Residual Heat Removal (RHR) Pumps: Not running. SELD:T the statement that; describes the effect that the Low Pressure core Spra/ (LPCS) pump motor breaker falling to close would have on the Automatic Depressurization System (ADS).
- a. ADS automatically actuates at this time. ,
- b. ADS will NOT automatically actuate.
- c. ADS will actuate 105 seconds after the ADS "A" manual initiation pushbuttons are depressed.
- d. ADS will actuate when the ADS "B" manual initiation pushbuttons are depressed.
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, SENIOR REACTOR GPERATOR PCg3 21 !
i QUESTION: 027 (1.00) Given the following plant conditions: i PLANT CONDITIONS The plant is operating at rated conditions. The High Pressure Core Spray system (HPCS) failed its full-flow surveillance test (could NOT achieve rated flow) . RHR pump "A" is tagged out of service for maintenance scheduled to- , l last for two more days.\ '" j eg r o ( t , tM i (O r h SELECT the a$ tion required for these plant conditions. Applicable sections l cf PAP-606, " Condition Reports and Immediate Notifications" and Technical Sp;cifications are provided as Attachments 4 and 17. !
- a. A 14 day LCO to recover HPCS is initiated, a one-hour report is made to the NRC.
j
- b. A 14 day LCO to recover HPCS is initiated, a four-hour report in made to the NRC.
c. A 6 hour LCO to hot shutdown and cold shutdown in the following 24 hours is initiated, a one-hour report is made to the NRC. '
- d. A 6 hour LCO to hot shutdown and cold shutdown in the following 24 hours is initiated, a four-hour report is made to the NRC. l l
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I SENIOR REACTOR OPERATOR P;g3 22 . I i QUESTION: 028 (2.00) J
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MATCH each of the Reactor Protection System (RPS) A with its Technical Specifications basis in column B. trip parameters in column may be used once, more than once, or not at all. ONLY aItems SINGLE in column B answer is uccd for each item in Column A. COLUMN A COLUMN B (RPS TRIP PARAMETER)
.................... (TECHNICAL SPECIFICATION BASES) !
- a. Main steam line high 1. Anticipates a pressure and radiation, neutron flux increase and i provides pressure and fuel
- b. Main steam line isolation thermal / hydraulic safety limit valve closure. protection. ,
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- c. Reactor vessel high water level,
- 2. Compensates for reactivity addition and subsequent power ,
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- d. Reactor vessel steam dome rise due to cold water '
addition. high pressure. I
- 3. Reduces the continued failure f of fuel cladding.
- 4. Compensates for reactivity addition from void collapse ,
and resulting power rise.
- 5. Provides protection from exceeding the low pressure /
low flow thermal power safety > limit.
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SENIOR REACTOR OPERATOR Pago 23 QUESTION: 029 (1.00) SELECT the condition that would cause a reactor scram,
- a. Turbine Stop Valves (TSVs) "A" and "B" are closed and turbine first stage shell pressure is 50% power.
- b. Turbine Control Valves (TCVs) "A" and "C" emergency trip supply (ETS) pressure is 650 psig and first stage shell pressure is 25% ,
power.
- c. Turbine Control Valves (TCVs) "A" and "D" emergency trip supply (ETS) pressure is 230 psig and first stage shell pressure is 45%
power.
- 1. Turbine Stop Valves (TSVs) "B" and "C" are closed and turbine first stage shell pressure is 40% power.
QUESTION: 030 (1.00) - SELECT the_ statement that describes a condition in which the Intermediate R2nge Monitoring (IRM) circuitry is NOT functioning-correc21y. The reactor ande switch is in "STARTUP/ HOT STANDBY."
- a. IRM "A" indicates 90/125 scale on range 7 and '40 rod block or scram signal is present.
- b. IRM "B" indicates 122/125 scale on range 6, a rod block is present, and IRM upscale trip light is lit.
- c. IRMs "D" and "A" both fail full upscale and a reactor scram occurs.
- d. IRM "C" mode / test switch is placed to " STANDBY" and a rod block and half-scram occurs.
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SENIOR ~ REACTOR 5PERATOR P;ga 24 QUESTION: 031 (1.00) , Given the following plant conditions: ,' PLAINT CONDITIONS' The reactor mode switch position: "STARTUP/ HOT STANDBY." Reactor power is midscale on range 4 of the Intermediate Range Monitors (IRM). Source Range Monitor (SRM) D is failed upscale and is bypassed. All IRMs are selected to range 4. SELECT the statement that describes an operating condition that will cause th3 Rod Control and Information System (RC&IS) to generate a rod block. a. The SRM channel A indicates 7 E 4 cps and its detector is fully inserted. b. ThJ SRM channel C indicates 50 cps and its detector is partially withdrawn. c. The IRM channel B indicates 30/125 scale and its detector is fully: inserted. d. The fullyIRM channel F indicates 8/125 scale and its detector is not inserted. l 1
SENIOR REACTOR OPERATOR POg3 25 l l QUESTION: 032 (1.00) SELECT the statement that describes the Average Power Range Monitoring cyctem (APRM) input to the recirculation flow control system. ,
- a. APRM "E" output is compared with actual core flow to generate the l loop flow demand signal.
- b. APRM "A" output is compared to actual core flow signal to shift the loop flow. controller to manual if not in the allowed area of ,
the power-flow operating map. l
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- c. APRM "A" output is compared to the load-following circuit's load demand error signal to generate the flux error signal, matching i reactor power to generator load.
output is hompared to the output of the flux controller
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- d. APRM "E" to generate the flow demand signal. i QUESTION: 033 (1.00) ,
SELECT the statement that describes an operational condition when the R3Gctor Core Isolation Cooling (RCIC) system would FAIL toeinitiate. 9t o.ht ; n to
- a. RPV water level is +18 inches and the B21-N692A level instrument ,
is failed upscale.
- b. RPV water level is +20 inches and the 125 vde bus ED1B is deenergized.
- c. RPV water level is +105 inches and the 125 vdc bus EDIA is deenergized.
- d. RPV water level is +110 inches and the reference leg for the B21-N692A and B21-N692E level instruments is ruptured.
i i i SENIOR REACTOR OPERATOR PO93 26 { i i QUESTION: 034 (1.00). i SELECT the statement that describes the sequence of actions riecessary to I rocet and open the Reactor Core Isolation Cooling (RCIC) turbine trip and throttle (T&T) valve after it trips due to overspeed protection. l
- a. Take the T&T valve control switch to close until the green light is on and the red light is off, reopen the T&T valve with the ,
control switch, reset the mechanical linkage locally.
- b. Reset the T&T valve mechanical linkage locally, take the T&T valve ;
control switch to close until the green light is on and the red light is off, reopen the T&T valve with the control switch. l l
- c. Take the T&T valve control switch to close until the green light is on and the red light is off, reopen the T&T valve with the control switch.
- d. Take the T&T valve control switch to close until the green light ;
is on and the red light is off, reset the T&T valve mechanicel linkage locally, reopen the T&T valve with the control switch. ! \ QUESTION: 035 (1.00) SELECT the statement that describes the Automatic Depressurization System (ADS) interrelationship to the DC electrical system. i
- a. All 8 ADS valves will open-if BOTH subchannels in EITHER the A ~
l < channel logic powered from division I 125 Vdc OR the B channel l logic powered from division II 125 Vdc are tripped.
- b. All 8 ADS valves will open if ONE of the subchannels in EITHER the A channel logic powered from division I 125 vdc OR the B channel l
logic powered from division II 125 vde are tripped..
- c. Only 4 ADS valves will open if division II 125 vde power is I
available and ONE of the subchannels in the B channel logic is ' tripped.
- d. Only 4 ADS valves will open if division I 125 vde power is available and BOTH of the subchannels in the A channel logic arc tripped.
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i SENIOR REACTOR CPERATOR P 93 27 [ QUESTION: 036 (1.00) I SELECT the immediate action required if the fuel pool cooling and cleanup (FPCC) return line to the fuel pool ruptures causing tP. s&weipt of the following annunciators. Fuel is being offloaded from the re, actor. ANNUNCIATORS CONTAINMENT FUEL STRG POOL LEVEL LO LO CONTAINMENT FUEL STRG POOL LEVEL LOW FPCC SURGE TANK A LEVEL HI/LO FPCC SURGE TANK B LEVEL HI/LO 6 a. Irrespective of location of the refueling platform, lower any loaded bundles until the slack cable light illuminates. P
- b. Hove any bundles located in the inclined fuel transfer machine and [
place them in a fuel storage rack in the spent fuel pool. 1 c. Immediately stop all fuel movement and evacuate the containment and drywell.
- d. Lower any loaded fuel bundle into any vacant fuel cell in the reactor vessel. l QUESTION: 037 (1.00)
SELECT the statement that describes a situation where ALL power would be l intorrupted further to the motion mainhoist. of the fuel hoist on the refueling platform preventing ANY *
- a. A fuel bundle is raised until the overhoist light is illuminated. I
- b. A fuel bundle !s raised until tha grapple normal up light is illuminated, c.
The refuel platform the traverse is in pushbutton, system stop the predetermined " red zone"' lighting
- d. A fuel bundle is being removed from the inclined fuel transfer carriage while a control rod is being withdrawn.
P;g3 28 SENIOR P.EACTOR OPERATOR
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i QUESTION: 038 (1.00) ; Given the following plant conditionst PLANT CONDITIONS , l Reactor powert 65%
"A" MSL radiation monitor (K610A) mode switch is in " STANDBY." l l
SELECT the statement that describes the effect of "B" Main Steam Line (MSL) radiation monitor (K610B)-failing upscale. ; l
- a. Only the inboard MSIVs isolate, the reactor scrams on.MSIV closure.
- b. The inboard and' outboard MSIVs isolate, the reactor scrams on main :
steam line (MSL) high radiation. ,
- c. The "1/2 SCRAM A/C" alarms, the inboard and outboard MSIV channel f "A" pilot solenoids deenergize, only the inboard MSL drains ,
isolate.
- d. The "1/2 SCRAM A/C" alarms, the inboard and outboard MSIV channel "B" pilot solenoids deenergize, no valve isolations occur. :
L t k l l 4
a POg3 29 SENIOR REACTOR CPERATOR QUESTION: 039 (1.00) SELECT the statement that describes how the Low-Low-Set (LLS) function of tho Safety Relief Valvos (SRVs) automatically operates following a main ctcam line isolation at rated power conditions.
- a. 1B21-F051C opens at 1103 psig arming LLS, which opens five LLS SRVs until 1113 psig, 1B21-F051D cycles between 1073 psig and 926 psig.
- b. 1B21-F051C and D open at 1103 psig arming LLS, four LLS SRVs open at 1113 psig until 936 psig, 1B21-F051C and D cycle between 1073 psig and 926 psig.
- c. 1B21-F051D opens at 1103 psig arming LLS, five LLS SRVs open at 1113 psig until 946 psig, then four close. 1B21-F051D and C cycle between 1033 psig and 976 psig.
- d. 1B21-F051D opens at 1103 psig arming LLS which opens 1B21-F051C, four LLS SRVs open at'1113 psig until 946 psig, 1821-F051C closes at 936 psig, 1821-F051D cycles between 1033 psig and 926 psig.
i
[ I SENICR REACTOR CPERATOR Paga 30 ; i P I QUESTION 040 (1.00) ; i Giv0n the following plant conditions f PLANT CONDITIONS i Reactor povert 45% Generator loadt 560 MWe r Recirculation flow controit Loop manual. I SELECT the plant response to a continuous runback of the load set demand
.cignal to zero in the Electro-Hydraulic Control (EHC) system. r
- a. The turbine control valves (TCVs) throttle closed, the bypass valves (BPVs) remain closed, reactor pressure increases, the ,
reactor scrams on high pressure or high' neutron flux. -
- b. The bypass valves (BPVs) throttle open, reactor pressure l decreases, the MSIVs isolate on low steam line pressure, the reactor scrams on the MSIV closure.
- c. The turbine control valves (TCVs) throttle closed, the bypass valves (BPVs) throttle open to compensate, once the BPVs are fully open reactor pressure increases causing a scram on high pressure or neutron flux.
- d. The turbine control valves (TCVs) throttle closed, the bypass valves - (BPVs) throttle open, reactor pressure remains fairly i constant, reactor power increases slightly due to reduced l i
feedwater heating. . L 3 4 G S
?
. . - . . - - . - ., =,. ..,.,;. - , , - . . , , - - - ,
e
.8ENIOR' REACTOR OPERATOR Pega 31 l l
V
)
. QUESTION: 041 (1.00) ;
-Given-the following plant conditions f PLANT. CONDITIONS
+
Reactor power: 100%. ; Generator load: 1245 MWe. i Standby pressure regulator: In " TEST." ! 1 SELECT the action required if the Electro-Hydraulic Control f.EHC) turbine e
, inlet pressure sensor fails HIGH. i
- a. control reactor pressure by using the bypass valve jack and {
commence'a power reduction, l
- b. Match steam flow to reactor power by reducing the maximum combined' '
I flow setpoint.
- c. Manually scram the reactor and maintain cooldown rate lese chan 100T/hr by closing the MSIVs. :
- d. Manually scram the reactor and manually control reactor pressure ,
below 1065 psig using the bypass valve jack. .; QUESTION: 042 (1.00) SELECT the statement that describes the relationship between the condensate Syctem and the Control Rod Drive Hydraulic (CRDH) System. ;
- a. CRDH normal water supply is from the condensate booster pump (CBP) ;
, suction, its alternate supply is from the condensate storage Tank (CST).
- b. CRDH normal water supply is from the hotwell pump suction line, .
its alternate supply is from the Condensate Storage Tank (CST).
- c. CRDH normal water supply is from the hat surge tank, its alternate
! supply is from the domineralized water header.
- d. CRDH normal water supply is from the hot surge tank, its alternate
! supply is from the Condensate' Storage Tank (CST). ; j - t
I SENIOR REACTOR OPERATOR Pag 3 32 QUESTION: 043 (1.00) ' SELECT the statement that describes the operator action required when the H:t Surgein roculting Tank the level controller following plant malfunctions conditions. causing its level to increase PLANT CONDITIONS I
" HOT SURGE TANK LEVEL HI" annunciator alarms.
Hot surge tank level is off-scale high.
- a. Immediately reduce core flow to minimum.
- b. Manually open the reactor feed booster pump minimum flow valve, t
- c. Immediately trip the main turbine, d.
Trip all hotwell pumps and condensate booster pumps. QUESTION: 044 (1.00) : SELECT the operational conditions required that will automatically start -tha Motor-driven Feedwater Pump (MFP) when the "A" Reactor Feedwater Pump i Turbine (RFPT) trips, leaving only the "B" RFPT in service.
- a. MFP control switch position:
"A" RFP hydraulic trip header pressure: AUTO.
90 prig. MFP suction valve position: Closed. l
- b. MFP control switch position:
"A" RFP hydraulic trip header pressure OFF.
25 psig. MFP suction valve position: Open,
- c. MFP control switch position OFF "A" RFP hydraulic trip header pressure: '
65 psig MFP suction-valve position: Closed.
- d. MFP control switch position:
"A" RFP hydraulic-trip header pressure: AUTO 65 psig MFP suction valve position: Open.
I
c l SENIOR, REACTOR CPERATOR - P;ga 33 l [ QUESTION: 045 (1.00) SELECT the statement that identifies the limit and the corresponding reason t fcr the minimum lube oil temperature for starting a Reactor Foodwater Pump Turbine (RTPT).
- a. 70 F, to ensure the oil is fluid enough to prevent excensive- k vibration in the turbine when it is rolled.
- b. 70 F, to ensure the oil is fluid enough to prevent tripping the oil pumps on excessive current.
1
- c. 90 F, to ensure the oil is viscous enough to build up an " oil wedge" if the RFPT trips and engages on the turning gear. ,
- d. 90 F, to ensure the oil will reach its operating temperature' i within 15 minutes of rolling the RFPT.
QUESTION: 046 (1.00) , SELECT the signal that providos an interlock to ensure adequate net pccitive suction head for the components of the recirculation system.
- a. Differential temperature between the recirculation loop suctions,
- b. Hydraulic Power Unit pump discharge pressure, i
- c. Differential temperature between the Recirculation Pump suction and reactor steam dome. *
- d. Reactor water level (level 4).
; SENICR REACT;R OPERATOR !
P;ga 34 i QUESTION: 047 (1. 00) i Given the following initial plant conditions: PLANT CONDITIONS Reactor power: 65% Reector reedwater Pump Turbine (RTPT) "A" and "B" controllers are in automatic. : SELECT the"A" statement that describes the plant response to the Main Ster.m Line (MSL) flow detector failing downscale. i i
- a. Reactor water level will increase initially, then return to the tape-set value.
- b. Reactor water level will decrease initially, then return to the '
tape-set value. c. Reactor level higherwaterthanlevel the will increase tape-set until it stabilizes at some value, d.
.Reactor level lowerwater thanlevel will decrease the tape-set value,until it stabilizes at some ii 4
i i i.
n giSENIOR REACTOR OPERATOR P;gD 35 l
'l LQUESTION: 048 (1. 00) ;
Given the following initial plant conditions: PLANT CONDITIONS ' ' 13.8 KV Bus L10 is powered from startup transformer 200-PY-B : HPCS diesel generator is tagged out of service for maintenance. +
)
t SELECT the statement that describes the expected response of the AC clcctrical distribution system following a main turbine trip due to a main. g:nerator differential current lockout trip. Figure R10-2A, "4.16 KV and Ab;ve" is provided as Attachment 6.
- a. The Lil and L12 buses will NOT automatically transfer to the L10 ;
bus, but can be manually transferred,
- b. The L11 and L12 buses will automatically transfer to the L10 bus,
- c. The EH13 bus will be deenergized since the HPCS diesel generator is tagged out.
- d. Bus L12 will NOT transfer to the L10 bus when the L10 bus in powered'from startup transformer 200 PYB. .
L t I k
SENIOR REACTOR OPERATOR Pag 3 36 QUESTION: 049 (1.00) Given the following initial plant conditions: PLANT CONDITIONS , Reactor mode switch is in " SHUTDOWN." Reactor coolant temperature is 250 F. l Unit 2 Div.1 battery (2R42-S002) is inoperabic. i SELECT the statement that describes the action required for a fire in the Unit 1.Div. 1-(1R42-S002) battery room. Technical' Specifications are pr;vided as Attachments 17.
- a. -Restore at least one of the two batteries to operable within 2 hours or be in cold shutdown within the following 36 hours.. i
- b. Restore at least one of the two batteries to operable within 2 ,
hours or be in cold shutdown within the following 24 hours,
- c. Suspend operations with the potential of draining the reactor -
vessel.
- d. Declare HPCS inoperable, and take the actions of 3.5.1 (be in cold shutdown within 24 hours). .
i e I + , _. - , - -
SENIOR REACTOR OPERATOR pag 3 37 I QUESTION: 050 (1.00) { i SELECT the statement that describes an operational condition that will . cause the division III Diesel Generator (D/G)_ to automatically trip while l it is operating. ,
- a. RPV water level: +10 inches. >
Drywell pressure: . 2.9 psig. Engine jacket water temperature: 215 F.
- b. RPV water level: +165 inches.
Drywell pressure: 1.5 psig. Engine crankcase pressure: 5 psig,
- c. RPV water level: +165 inches.
Drywell pressure: 2 psig, t Engine bearing oil pressure: 14 psig,
- d. RPV water level: +32 inches. -
Drywell prescure: 0.5 psig. Engine speed: 450 rpm. QUESTION: 051- (1.00). SELECT the statement that describes the correct Diesel Generator (D/G) control manipulation when it is being operated in parallel with the grid.
'1
- a. Generator VARs are increased to adjust the power factor by taking ]
the D/G voltage regulator switch to " RAISE."
- b. Generator VARs are decreased to adjust power factor by taking the~
D/G governor switch to " LOWER."
~
- c. Generator kW.is increased to prevent a reverse power trip by taking the D/G voltage regulator switch to " INCREASE."
,d . Generator kW are decreased to prevent overloading the diesel by-taking the D/G governor switch to " RAISE."
l l
L Pago 38 i SENIOR REACTOR OPERATOR : i QUESTION: 052 (1.00) i
~Giv0n the following ' initial Offgas system alignment:
INITIAL OFFGAS. LINEUP Charcoal adsorber control switch: " BYPASS". -
- Charcoal adsorber bypass valve (1N64-F045): Open. ->
Adsorber inlet and outlet valves Closed. j (IN64-F051A-D and 1N64-F053A/B): In service. !
"A" Steam Jet Air Ejector (SJAE):
SELECT the expected realignment of the offgas system valves when the t following radiation monitor indications occur. RADIATION MONITOR INDICATIONS The'K601A post-treat radiation monitor is alarming and its "HI" indicator. light is illuminated. The K601B post-treat radiation monitor is alarming and its "DOWNSCALE/INOP" indicator light is illuminated.
- a. Charcoal adsorber bypass valve (1N64-F045): Open.
Adsorber inlet and outlet valves: Closed. (1N64-F051A-D and 1N64-F053 A/B) Open. Offgas discharge valve (1N64-F632):
- b. Charcoal.adsorber bypass valve (IN64-F045): Closed. *
' Adsorber inlet and outlet valves: Open.
(IN64-F051A-D and 1N64-F053A/B) I Offgas discharge valve (1N64-F632): Open.
- c. Charcoal adsorber bypass valve (1N64-F045): Closed.
Adsorber inlet and outlet valves: Open. (1N64-F051A-D and 1N64-F053A/B) Closed.
'7 Offgas discharge valve (1N64-F632):
- d. . Charcoal adsorber bypass valve (IN64-F045): Open. '
Adsorber inlet and outlet valves: Closed. (IN64-F051A-D and 1N64-F053A/B) Closed. Offgas discharge valve (1N64-F632): I f i l 4 -
SENICR REACTOR OPERATOR I Pcg3 39 ; 1 QUESTION: 053 (1.00) SELECT the condition of the containment ventilation exhaust radiation m:nitors (K609A-D) purgs system inboard that will cause an isolation of the containment vent valvas. and '
- a. K609A is alarming "HIGH-HIGH" and K609C is alarming "INOP."
- b. K609B is alarming "INOp" and K609C is alarming "HIGH-HIGH."
c. K609D is alarning "DOWNSCALE" and K60')C is alarming "HIGH-HIGH."
- d. K609A is alarming "HIGH" and K609D are alarming "DOWNSCALE."
l l QUESTION: 054 (1.00) Given the following plant conditionst PLANT CONDITIONS A loss of all offsite power (LOOP) has occurred. All Diesel Generators (D/Gs) are running and tied to their buses. SELECT the statement that describes the expected alignment of the Control Room Heating, Ventilation, and Cooling (HVAC) system. .igure " M25/26-lC,
" Control Room HVAC and Emergency Recirculation Systems" is provided as Attnchment 7.
- a. The supply fan and the return fan are running.-
- b. The supply fan is running and the return fan is stopped,
- c. The supply fan and the return fan are stopped.
- d. The supply fan is stopped and the return fan is running, i
Es e L ~ SENIOR; REACTOR OPERATOR. Pags 40 , y QUESTION: 05!) (1.00)-
. SELECT the operator-action required to be taken by. procedure if the "A" W 1rGcirculation pump trips while operating at. rated power. Plant; conditions
- following the recirculation pump trip are listed below. Figure A006,
" Power to' Flow Operating Map," and ONI-B33-2,. Attachment 1, " Thermal-Power VOrsus. Core Flow' are provided as Attachment 8.
PLANT CONDITIONS'AFTER THE "A" RECIRCULATION PUMP TRIPS . Reactor power: 1 45% with 6%-bandwidth oscillations. 42 Mlbm/ hour Core flow: j 1 a.- Arm and depress the RPS manual scram pushbuttons. I b' . No inmediate operator actions are required. k
- c. Insert control rods in-reverse sequence'to lower reactor power below the limits of ONI-B33-2, Attachment 1.
- d. Insert the CRI.M rods to lower reactor power below the limits of. 4 i
ONI-B33-2', Attachment 1. .i
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4 __________i_i____
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-SENIOR REACTOR. OPERATOR Pago.41 f' . QUESTION: '056 (1.00),
Tho unit.is'in Operational Condition 2 during a reactor atartup when the "A"~ Recirculation Pump trips. Ten minutes later the "B" Recirculation Pump-trips.
~
SELECT the operator action REQUIRED fer the following plant conditions.. Th3 appropriate sections of' Technical Specifications and Figure A006,
" Power to Flow operating Map" are provided as Attachments 8 and 17.
PLANT CONDITIONS AFTER "B" PUMP TRIPS i i Reactor. power: 8%- Core' flow: 20 Mlbm/ hour
.a. Arm and depress the reactor manual scram pushbuttons immediately.. i
- b. Be in Hot Shutdown within six hours, c..
Be in Hot Shutdown within 12 hours. d. Reduce reactor power to within the~ limits of-figure 3.4.1.1-l' within two hours by inserting CRAM rods. o l t i t i (
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- k. SENIOR REACTOR' OPERATOR' e -
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sjQdESTION: 057' (1.00) r' l fde . . E! Givan the following; initial plant conditions: l t iINITIAL PLANT CONDITIONS' , Generator load: 350 MWe 32% Reactor. power: LSELECT1the statement that describes the expected; sequence of events with'no- , opbrator actions to-increasing pressure conditionsEin the main condenser. . 1 a.- The main turbine trips causing a reactor scram, the bypass valves a
,will close- the MSIVs will isolate.
o b. The main turbine triri causing a reactor scram, the.MSIVs Isolate ~, j
% -the; bypass valves close. i The main turbine trips,.the MSIVs isolate. causing a reactor-scram,
~
n c. the bypass valves-isolate. Ld. .The- main- turbine. trips,Ethe bypass -valves close, the'MSIVs isolate . causing a ree tor scram.
?
1 LQUESTION: 058 -(1. 00) .
; SELECT the~ statement =thatsdescribes the limit;andEcorresponding reason for
< Lthi: number ofLaanual diesel generator start attempts from'the. main control-t-roomLduringLafstation blackout.
I
'a.: Two attempts are allowed.to prevent damaging the journal. bearings'
" .during:,the unsuccessful starts.
t b.:?One attempt is allowed to prevent depleting the air. start I - receivers.-
*. 'I'c. Two. attempts are allowed to prevent depleting the air start. ..
receivers.- 4
!d. One attempt is allowed to prevent damaging'the journal bearings during the. unsuccessful starts.
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SENIOR REACTOR OPERATOR 1 Paga 43 j l
)
1 QUESTION: 059 (1.00)l l SELECT the statement that describes the consequences of failing to. complete-tho-immediate action of-transferring Reactor Feedwater Pump Turbine (RTPT) control from automatic to manual for the following plant conditions.. -i
-PLANT CONDITIONS DC bus D-1-A is faulted and doenergized.
Reacto( power: 100%. Lo d ti i s lidt (i t d. ,
- a. RPV water level continues to increase causing a reactor scram and damaging the main turbine because the A RFPT and main turbine will; '
not trip on high level.
- b. RPV water level increases until the RFPTs trip on high level,'but~
the reactor scram will not occur due to the loss of the DC bus.
- c. RPV vater. level' increases until the RFPTs trip on high level, the reactor will. scram when water level decreases to.t,he level 3 due I to loss of feedwater. 3
- d. 'RPV level will increase and stabilize at a higher level when level error' overrides the flow error signal on the "A" RFPT. .
QUESTION: 060 (1.00) . SELECTLthe statement that describes a plant condition that would cause the mnin turbine generator to automatscally trip. .
- a. Vibration is 10 mils for 2 minutes.
- b. Condenser pressure is 5.2. psia and generator load is 500 tNe. !
t
- c. ,RPV level is +220 inches. j
- d. Exhaust hood temperature is 145F.
t P t i
.f Paga 44 (SENIOR l REACTOR OPERATOR' s.
p QUESTION:n061 (1.00) x.
. SELECT >the operator-action that will reestablish control' Rod Drive .
~
LHydraulic-(CRDH) drive differential pressure.following a high drywell: Lprsssure scram that cannot be reset. - CRDH pump "A".is, running, ,
'a. ' Bypass the CRD pump sucticn filters. }
- b. Place the flow control valve in " MANUAL" and open it. ,
- c. Fully,close the drive water pressure control. valve.~ ,
- d. lTake the high drywell pressure scram bypass switches to " BYPASS" .
and reset the scram and alternate rod insertion (ARI). i 4 QUESTION:: 062 (1.00) L{
' SELECT the statement that describes the plant response as RPV water level increases-to Leve178 setpoint. High Pressure Core Sprayo(HPCL) and the-Motor Feedwater Pump (MFP) were manually. started during a . loss of: feedwater
- flow' transient:and recovered level.
- a. -The;HPCS pump will trip, the MFP will operate until manually; '
secured. t
- b. . :The HPCS injection valve will close and the MFP will trip.
4 _c. The HPCS pump and MFP-will trip. ; d .- ThelHPCS injection valve will'close, the MFP will operate _until
-g-
; manually secured.
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SENIOR REACTOR'UPERATOR ,
'Pago 45 QUESTION: 063'- .(1.00) 7 ' SELECT the statement that describes a condition that allows energizing! stub bus-XH12 during a large break a Loss Of Coolant Accident (LOCA) condition.
The caution note from PEI-D23-3, Section 5.0, "Special-Operations" is reproduced below. co6************************************************************************ CAUTION _ Before energizing the stub bus in the following steps, verify RPV' level can be determined, and no ECCS pumps associated with the respective stub bus are providing adequate core cooling.
~
Subsequent to the stub bus'energization event, when any of the ECCS pumps associated with the bypassed stub bus are providing a adequate core cooling, ' then open the ISOLATING BRKR EH1116(1214) and place the BUS XH11(12) LOCA BYPASS keylock switch in NORM. .o ooo***************************************************************t******** a. The RPVLow Pressure water level isCore Spray (LPCS)
-12 inches pump is injecting at 6500 gpm, and stable,
- b. The Residual Heat Removal (RHR) pump RPV: water level is +4' "C" is injecting.at 5000 gpm,.
inches and stable.
;c.
The Low Pressure Core Spray (LPCS) pump is injecting at 3500.gpm, RPV water level is +5 inches and increasing.
- d. The Residual Heat Removal (RHR) pump "B" is injecting at 7000.gpm, RPV water level is -5 inches and increasing.
4 e
- p. SENIOR /REACTvu OPERATOR ;ptg3 46-
-i
, QUESTION:'064 ( 1. 00):
) SELECT the condition that1 requires entry into'the plant emergency Lin tructions.-- ^ ~
- a. RPV water level is +182 inches .
- b. An instrument: volume high level signal is present, 25 rods failed to insert, reactor power is 75/123 scale on IRM range 4. .
- c. Drywell-temperature Is 130 F.
- d. Suppression pool; level is 19.5 feet.
V
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a , i SENIOR 4 REACTOR OPERATOR' Pago 47- l
?
N i QUESTION:2 065 -(1400) .i I Given the following plant conditions: \
- PLANTm CONDITIONS' !
)! Containment temperature: -192 F.
Containment pressure: . 1.5 psig. Drywell temperature: 245 F. Drywell pressure:- 4.1 psig. _
. RHR "A" pump is inoperable. .
RCIC;is. injecting and. maintaining RPV water level at'+120 inches., ,! i
'SELECTLthe action-required:after a scram has occurred on high drywell!
l .
' prassure.LThe applicable sections of PEI-D23-1, " Containment Temperature 1 iniprovided Control" as .and PEI-D23-2,:
Attachment "Drywell and Containment-Pressure Control" are-
=10. l
, a. Initiate containment sprays.: '
.! : i
- b. ; Emergency ~depressurize the reactor vessel ~.' '
l< c. Operate the backup Hydrogen purge mode of combustible gas control.
. d. Vent
~
the~ containment with the fuel ~ pool cooling and' cleanup system. .(
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Pag 3:48 ! fiSENIOR REACTOR OPERATOR \/ r J 1 -l QUESTION: 066 '(1.00) SELECT the. operational situstion that could cause the following! indications-while' operating ~at 85% power, 1 cINDICATIONS Annunciators alarming APRM A/E UPSCALE ROD BLOCK APRM UPSCALE RPV: pressure > increasing.
- a. -Maximum combined flow limit fails high. P
- b. Pressure 1 regulator fails open.
- c. Pressure regulator fails closed.
i
- d. Bypass va).ve fails open.
.l
; QUESTION: 067 .(1.00)
~
Given the following plant conditions: PLANT CONDITIONS . 3
- Reactor power: 4 4. -
Main steam lines are isolated. 7
- Suppression, pool-cooling is operatingL(both loops).
' SELECT the MAXIKUM Suppression Pool' temperature that can exist BEFORE Standby Liquid control (SLC)' MUST be injected. PEI-B13, Figure 1, " Boron
- Injection Initiation Temperature" is.provided.as' Attachment 9. , ,
.a.. "121 F. .I
, J
- b. 120 F. .L
- c. '110 F
- d. 109 F.
I ESENIOR. REACTOR OPERATOR Paga 49:
, -QUESTION: 068 (1.00) I SELECTithe reason.that the green LEDs on the P680 Rod Display. Module for?
ccch! control rod:are checked following a failure to scram. a.- To confirm'that the scram pilot valve solenoids on each HCU are deenergized, b. To confirm-that the scram pilot air header'is depressurized. c '. To confirm thatieach HCU accumulator is. pressurized. ! d. To confirm open. that the scram inlet and outlet valves on each HCU are- f i
' QUESTION:l069_
(1.00)
. SELECT (room the-operator if capable-of beingaction that is' required prior to' evacuating the control performed. l
- a. r Place the reactor mode switch in " SHUTDOWN."
t b.
-Place all MSIV control' switches to "CLOSE."
c. Transfer-station loads from the auxiliary to startup transformer, d.
., Start the division 1 and 2 diesel _ generators.
b (
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.l , SENIOR.lREACTORtOPERATOR Pags 50 '
9 h i QUESTION: 070-- (1. 0 0)- f
' SELECTL the statement- that identifies a notification requirement- during a Sito Area Emergency due to a steam line break in the turbine building that .
- cunnot be isolated.
- a. Notify the NRC within one hour. t
- b. Notify'Ashtabula county' officials within one hour. j
- c. Notify.the State of Ohio within 30 minutes.
- d. Notify the- VP Nuclear Group within one hour. d QUESTION: 071 (1.00)
, SELECT the action that is requ' ired to be performed immediately following a total loss of Nuclear, Closed Cooling Water. [
.a. . Trip the Recirculation pumps immediately. ,
- b. Trip'the Recirculation pumps.within-three to five minutes.
t
- c. Trip the Recirculation pumps'if a high temperature alarm ir, received.
- d. Shift Recirculation pumps from fant to slow speed.
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' SENIOR REACTOR OPERATOR >
' Pagt 51 I4
' QUESTION: 072 (1.00)
(Givenithe'following. plant' conditions: 2 PLANT; CONDITIONS Annunciators::
. SERV AIR COMP TRBL TBCC HX' OUTLET TEMP HIGH NCC HX OUTLET TEMP HIGH.
OUTLET TEMP:HIGH (stator cooling water)
'RWCU isolation valve, 1G33-F004, isolates Loffgas glycol refrigeration compressors trip. ;
SELECT the malfunction that caused the conditions to occur.
-a.-- Loss of Turbine'Duilding' Closed Cooling' Water. +
- b. Loss of Service Water, "c. . Loss of Service:and/or Instrument Air. i i
c 'd, Loss of' Nuclear Closed Cooling Water.
- QUESTION: 073 '(1. 00) i
-i
-: SELECT the statement that describes a' condition requiring a fast-reactor cchutdown following a loss-of instrument air'while operating at 45%. power. ,
a._.The SCRAM VLV AIR HEADER PRESSURE LOW annunciator alarms. bi The! ROD DRIFT' annunciator (for-rod 22-11) and CRD MECHANISM TEMP- ) HIGH annunciator'(for rod 46-39) alarms.4
.c./-The ' driftsoutboard
~ closed.main steam _line isolation valve (MSIV) B21-F028A' i
~ >
d. The~ INST VOL NOT DRAINED annunciator alarms. f - 6 - -- -, -n e
-Pag 3 521 SENIOR' REACTOR OPERATOP. -QUESTION: 074 (1.00)
SELECT the statement that describes an approved alternate, method of decay hOtt removal that can be used 3 f the normal shutdown cooling lineup is lost and cannot be reestablished while in operational condition 4.
- a. Operate High Pressure Core Spray (HPCS) to circulate coolant between the suppression pool and the reactor vessel via the head vent.
- b. Operate-Low Pressure Core Spray (LPCS) to circulate the coolant between the suppression pool Ond the reactor vessel with LPCS via e a Safe *.y Relief Valve (SRV).
c.. Operate the condensate system to circulate water between the main condenser and the reactor vessel with the hotwell pumps via the Main Steam Isolation Valves (MSIVs).
- d. Operate Low Pressure Core Spray (LPCS) to maintain level while dumping to radwaste from Reactor Water Cleanup (RWCU).-
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[ SENIOR; REACTOR OPERATOR. ~ .P093 531 1 J i i QUESTION: 075 (1.00) , Givon the following initial plant conditions: INITI$L PLANT CONDITIONS
-Bottom head (drain temperature: 180 F.
;Recircl loop'"A" suction temperature: 188 F.
+235' inches. 3
~
T RPV' level: RPV head-vent: Open.- ! Residual Heat Removal (RHR) loop "A": Operating in= Shutdown- l
' Cooling.
SELECT the statement that describes the concern regarding the "A" RHR pump
; tripping while operating in shutdown cooling mode.
- a. In condition 4, one: loop of shutdown cooling must.be in service at all times, therefore this constitutes a Technical Specification violation.
- b. The coolant may heatup without adequate monitoring cap 2bility, Sand- ,
the plant may shift from condition 4 to condition-3.
. Oxygen may. enter the vessel through the open head vent causing c.
increased: stress cracking corrosion. d.: The fuel clad may undergo a zirc-water' reaction _ releasing hydrogen', i gas into the containment.
+
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! SENIOR;REACTORLOPERATOR. Pag 3 54-
, i i
(1.00)'
-QUESTION: 076 B i ..Given,the following' plant conditions:
. PLANT CONDITIONS ;
Reactor pressur.e is 855 psig. i
. Cbntrol rod 22-11-is at-position 00, its nitrogen accumulator has a. !
' cracked weld and is isolated for repair.
~ SELECT. the action required if the operating . Control Rod Drive > (CRD) pump Ltripsland a CRD pump CANNOT be restarted. The CRD LEVEL HI/ PRESS LOW >
!Gnnunciator is received-for.the=following' rods and the PPO has verified-
' accumulator pressure at-1350 psig for the alarming HCUs. ]
l
' Rod Position r
...____' ___ .... i 18-271 00 ,
38-23. 48- ,
.i
< a.- Immediatcly place the reactor mode switch to " SHUTDOWN.". q
- b.: IfLany otherfaccumulator becomes inoperable.for a withdrawn rod, Dimmediately place the' reactor mode switch in " SHUTDOWN."
j
. c; If any other accumulator becomes. inoperable for' a withdrawn rod,
~
start:a CRD pump within 20 minutes-or place the reactor mode 4 i switch-in " SHUTDOWN." d.- Start.a'CRD pump.within 20 minutes or place the-reactor mode switch in " SHUTDOWN."
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? SENIOR REACTOR OPERATOR
- Paga 55 qQUESTION: .'077 (1.00) 3 II
-t SELECT the statement'that describes the expected plant response'following'a. Ei rafueling1 accident where an exposed fuel bundle is dropped.and"is damaged.
Tha control room annunciation is listed below.
. CONTROL ROOM ANNUNCIATION CNTMT VENT EXH RAD HI. -
CNTMT VENT EXH RAD A/D HI HI/INOP
, CNTMT VENT EXH RAD B/C HI HI/INOP j[
- a. Vent: supply and exhaust dampers remain open, the supply and-exhaust fans trip on the high radiation signal. ,
- b. Vent supply and exhaust dampers isolate and the supply and exhaust.
~
fans trip when the dampers isolate. < l
- c. Vent exhaust dampers isolate and the. exhaust fans trip when the dampers isolate, the supply fans continue to operate.
- d. Vent supply dampers isolate and the supply fans trip on the high. '
radiation signal,'the exhaust fans continue to operate. 4
-i
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i SENIOR! REACTOR: OPERATOR Pcga 56 y 1
- l 1 -QUESTION: 078- (1.00)l /
o I :i
~
LGiven-:the following plant conditions:
. PLANT CON')ITIONS: _
RPV pressure: 80 psig., Dryvell pressure: 58 psig. : Drysell-temperature: 320 T.
<< 1 Containment pressure: 54 psig. .I Containment temperature: 135 F. !
Suppression' pool water level: 32 feet. ; SELECT.the action _ required and its corresponding reason following a-failure: toEscram and steam leakLin N:e drywell. Applicable sections of PEI-D23-2, "Drywell and Containment-Pressure Control" are provided as Attachment 10. 1
- i
- a. Vent the containment to ensure a controlled, monitored release 3 occurs rather.than an uncontrolled release from a failed 1 containment.
_l
- b. Initiate containment sprays.to reduce. containment ra'diation levels }
by removing fission. products from the containment atmosphere with , the " scrubbing action" of the sprays. L c. Initiate _the-backup hydrogen purge mode of the combustible-gas control: system toLensure a saturated drywell atmosphere can; be-obtained for-maximum pressure reduction when cooling is restored ~ l d. ' Emergency depressurize_the reactor to-ensure _that-containment.
- '. pressure will not exceed the containment design pressure ~ limit.
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SENIORLREACTOR OPERATOR _P393 57 ; QUESTION: 079 (1.00) l SELECT'the one statement below that identifies an action required, if any, ; following a reactor scram-due to-an isolation of all Main-Steam Isolation + valves (MSIVs) at rated power.if reactor pressure peaks at 1345 psig, i
- a. Notify the NRC within four hours,
- b. Do not restart the unit until authorized by the NRC. '
- c. No further actions required,-'the plant is~in hot shutdown.
- d. Notify the VP-Nuclear Group within - four hours, t
' QUESTION: 080 (1.00) ,
e i i SELECT-the statement that describes the Redundant Reactivity Control System. .!
'rteponse.that will occur 25 seconds after pressure exceeds 1083 psig following a failure to scram on a main turbine trip at-45% power.
r a. Feedwater are flow is reduced to zero unless the feedwater controllers in manua]. b.- I Alternate later. rod insertion will actuate and can be reset-30 seconds !
- c. 'The reactor recirculation pumps _will trip.
a,
- d. Standby Liquid Control will automatically initiate.
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-l SENIOR; REACTOR OPERATOR 'Pago 38:
l l 1 1
' QUESTION: 081- (1.00)- '
1 SELECTthestatement1thathescribesanoperatoractionrequiredforthe
'following plant conditions.
PIA 14T CONDITIONS -- React'or power: 75%. Suppression pool temperature: 111 F. 3
. Suppression pool level: 18 feet 2 inches. l SRV 1B21-F047B: Failed open. ;
a
- a. If the!SRV cannot be closed within'five minutes place the reactor mode switch in " SHUTDOWN." '
- b. Immediately place the reactor mode switch in " SHUTDOWN."
.c. Reduce suppression pool level to less than 18 feet.
Ed. If suppression pool tem earature exceeds 120 F, arm and depress.the. manual scram pushbuttons. v i L
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I.b - r Is PagGL59 SENIOR REACTOR OPERATOR
- 1
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-4
, QUESTION: 082, (1.00)-
'Givsn the1following plant conditions: ;
i
. PLANT. CONDITIONS
-;l Drywell temperature: 325 F.
Containment temperature: 220 F. .' l RPV pressure:- 100 psig.
'i C SELECT'the statement that describes an RPV water level instrument that may b3 used'for trending the reactor's water level. PEI-D23-3, Figures 1, 2b,
- )
2c,-2d, and 2e, are provided'as' Attachment 12.' l f'
- a. Wide-range. level indicates +9 -inches.
b..-Upset range level-indicates +190 inches. 4
.c.- Shutdownfrange indicates +200 inches.
- d. Narrow range indicates +170 inches. l
, a a
i f H
' QUESTION: 083 (1. 0 0)-
' SELECT the statement'that describes the reason emergency depressurization: :7
, .ic required ~if. containment: temperature cannot'be maintained less'than I 185 F.-
l 1
- a. MaintainiequipmentLoperability by reducing further energy addition. J from the' reactor to.the containment.. i
- b. Operability;of.the-safety relief valves (SRV) is in jeopardy land- '
[ it is: appropriate to emergency.depressurize before'they fail. ,
~
;The suppression pool may no longer be.able to absorb the energy 1, c.
from a loss of~ coolant accident.
.d. Containment liner failure may be-imminent and it is necessary to:
place the reactor'in a-low energy state. 1
' P ga 60 ' SENIOR REACTOR OPERATOR . QUESTION' 084 (1.00)
SELECT the statement that describes the reason reactor vessel water level-instruments may not be used for level-indication if the drywell temperature nOtr, the reference legs is greater than the RPV Sat'; ration Temperature.
- a. The differential pressure' transmitters are not environmentally qualified to operate at saturated temperature conditions.
- b. Drywell pressure is the same as reactor pressure providing a zero differential pressure (upscale level indication),
- c. Actual RPV water level may be lower than indicated water level.
- d. The density of the water in the reference leg is too low to provide a usable differential pressure to measure level.
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[ SENIOR' REACTOR OPERATOR '
-Paga.61, 7 i
' QUESTION:'0851 (1.00)
.Given the following plant conditions:
! PLANT CONDITIONS. '
Reactor pressuret 950 psig..
- Suppression-pool level: 26.5 feet. '
Suppression-pool temperature: 150 F. i SELECT the action; required and its corresponding reason. . PEI-G42,. Figures ,
.loond 2, " Heat Capacity Level ~ Limit" and " Suppression Pool Load Limit" are provided as Attachment 13.
- a. Emergency depressurize.the reactor:to prevent excessive- >
suppression pool dynamic loading due to SRV actuations,
- b. Emergency depressurize the. reactor to ensure the suppression pool i
can absorb 1the energy. released by a LOCA'. i
- c. . Reduce suppression pool level to.less than.23 ft..to prevent containmentLdesign~ temperature from being. exceeded if,an ADS
' blowdown.were to occur.
d. Reduce reactor pressure to less than 800 psig.toLprevent exceeding the-containment design temperature in the event of a Design Basis [ 4 o Loss,of Coolant Accident. L -I l- i .) l l' l L.- lt l~ r i t ll i i I 1 i 1 'I
P
' SENIOR REACTOR OPERATOR ~ Pags -
b
-QUESTION: 086 (1.00)
SELECT the_ statement ~that describes the reason for requiring suppression
- pool level to be greater than 5.25 feet prior to commencing emergency
'd3 pressurization.
E
.a. The' Safety Relief. Valves (SRVs) tailpipe would be uncovered and discharge directly to containment, the containment design pressure could be exceeded..
- b. The submergence of the Safety Relief Valves'(SRVs) is insufficient- t' for complete exhaust steam condensation and the containment design pressure could be exceeded, c.- The reduced mass of water that is cleared from the tailpipe when the Safety Relief Valves (SRVs) is opened would place excess'ive velocity loading on the tailpipe,
- d. To provide sufficient back pressure on the-Safety Relief Valves (SRVs) to ensure tie SRVs will reclose when the control switches are.placed to "CLOSE."
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i 1 SENIORLREACTOR OPERATOR- = Pags c3; f
'I
{ " . QUESTION: 087: (1.00) ! w
- SELECT the statement that describes the actions required if the following '
plant conditions occur while testing. safety relief-valves. .The applicable
- csctions of Technical Specifications are provided as Attachment-17.-
t PLANTECONDITIONS Reactor' mode switch position:
"STARTUP"
-Reactor coolant temperature: 540 F. .
Reactor' power: 2%. .}
' Suppression pools temperature: H97 F. "
-Suppression pool level: 17 feet 8 inches.
. Residual Heat Removal-(RHR) system loop "A" is in suppression pool-
' cooling mode.
- a. Be in hot shutdown within 12 hours and in cold shutdown within the- 3 following 24 hours, nb. Reduce suppression pool water temperature to less than'90 F within-12 hours'or be in cold shutdown within the'following 24 hours.
- c. . ' Raise suppression ~ pool level to.18 feet.Within one hour or.be in-hot shutdown within the next 12 hours. .i
- d. Reduce suppression pool level to 17 feet-6 inches within one hour-or be~in cold shutdown within 24 hours..
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ESENIORnREACTOR OPERATOR -Pags 64' i-
' QUESTION::0881 (1' c 00): -,
t i 1 SELECT the alternate injection system lineup addressed by'PEI-B13, " Reactor. Prcssure Vessel' Control" that could-be used in the event reactor water ! ilsvel CANNOT1be maintained greater than zero inches. 4
- a. Standby Liquid Control (SLC) transfer pump; injecting via the.
' reactor water cleanup (RWCU) return'line.
- b. Emsrgency' Service Water (ESW) injecting vianthe Residual Heat Removal'(RHR) shutdown cooling to feedwater line.
- c. Emergency Service Water (ESW) injecting via the High Pressure Core Spray-(HPCS) injection'line. I
- d. The-fire water system injecting via the Low Pressure Core Spral H (LPCS)-injection line. ]
" QUESTION: _089 (1.00) i
.1 SELECT the statement that describes a situation where adequate core.coolingc
'iolassured. -PEI-B13, Figure 4, " Minimum Alternate Flooding Pressure" is provided as Attachment 14.
k \ l
.a. Reactor power is 7%, two Safety Relief Valves (SRVs) are=open, RPV water level is -11 inches, reactor pressure is 450 psig, Reactor- ,
Core Isolation Cooling-(RCIC) is injecting. - l
- b. One safety-relief valve (SRV) is open~, RPV pressure is 100 psig, RPV level is unknown, Residual Heat Removal (RHR) loop "A" is injecting.
l
- c. Two Safety. Relief Valves (SRVs) are open, RPV water level is 1
~50-inches, RPV pressure is 40 psig, no' injection systems are 1 operating. 1
-d. One safety relief valve (SRV) is open, RPV pressure is 800 psig, RPV watar level is -135 inches, no injection is available.
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q SENIOR-REACTOR OPERATOR: PagoL65 t 4 5 QUESTION: 090 (1. 00)l SELECT ~the reason reactor power decreases'as RPV water level is' lowered.
< .during'and:ATWS.. )
e
- a. The single phase natural circulation flowpath through the steam-separators'is broken causing the remaining water inside the shroud. '
to boil and. totally' void.the. core region.
- b. .The driving head from theLdowncomer water level is-reduced,- 3 minimizing core flow, increasing the voids in the core region.-
- c. Carryunder' increases, increasing the preheating of the coolant, 1 '
reducing the core' inlet subcooling of the coolant.
- d. The total mass of coolant in the reactor vessel is' reduced, causing it to rapidly-void the entire core region.
M L i ;- ( L ! 1 l' L u I
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.Paga 66' i p(SENIOR 1 REACTOR OPERATORJ t
. QUESTION:,091 (1.00)
While: operating at 100% power,.the following-conditions occur:. i
~'
IPLANT CONDITIONS Annunciators . ANNULUS ' A DIFF PRESS LOW ANNULUSLB DIFF PRESS LOW
" Annulus differential pressure: Zero inches of water gage ,
? SELECT the potential adverse consequence that could cccur if this situation.
10 not' corrected.
- a. l Access to con'tainment could be limited due to the differential ,
pressure across the access doors.;
- 13. .The annulustregion could be overpressurized in the' event.of aj. Loss'
.of Coolant 1 Accident (LOCA).
c.- A ground level,.unmonitored releare could occur-in the' event of a.
. design basis. Loss of Coolant Accident (LOCA). ? ,
vi
.d.; Excessive load'would be placed on the containment. Vent.andEpurgeL system;due1to increased inleakage to the containment.
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LSENIOR-REhCTOR OPED.NTOR' Pagai671
-?
~l fQUESTION:. 092 (1.00) q
'GivGn the.following plant conditionst
; . PLANT-CONDITIONS
'The reactor failed to scram.
SLC was; initiated.
-l " Power Control Using RPV Level" (PEI-B13, Attachment 5) has been-entered.
SELECT the statement that describes when' Standby;LiquidcControl (SLC)'
" . ' linjection can be terminated AND. RPV water level 'can be raisea- irom the top L
.of active fuel to'+200 inches following a failure to shutdown. PEI-B13,
. Attachment 5 is' provided aus Attachment - 15. .
'a. Power is on'IRM range-4'and' decreasing,~all SRVs,ari Olosed.;
~
- b. The reactor engineer. confirms the reactor will remain shutdown' !
with the-present rod configuration.
.c.- The SLC storage tank level is 190 gpm, .andLthe SLC. pumps.have ->
. tripped.
- d. All control rods are fully' inserted, power,is'on IRM range 2 and decreasing. R o
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SENIOR REACTOR CP ltATOR PCg3 68 QUESTIO!!: 093 (1.00) SELECT the statement that describes an action required in order to reset a re3. tor scram during an ATWS.
- a. Place the LO POWER SET PT DIV 1 and 2 BV' ASS keylock switches to
" BYPASS."
- b. Place the INST VOL LEVEL HI SCRAM BYPASS keylock switches to
" BYPASS."
- c. Arm and depress the RRCS MANUAL ARI pushbuttons.
- d. Place the reactor mode switch to "RUN."
p [ i i 4
t SENIER REACTOR OPERATOR l PCg3 69 l l l QUESTION: 094 (1.00) ! i Given the following plant conditions: PLANT CONDITIONS Radiation levelst Containment: 50,000 R/ hour. ' Site boundary! 800 mren/ hour. j Offsite Dose Calculations: Whole body at site boundary: 1.5 rem. Thyroid at site boundary 8.5 rem. i Reactor shutdown timet one hour. t t SELECT the protective action recommendation required. EPI-BB, " Protective Actions and Guides" Attachments 2 and 3 are provided as Attachmonts 16.
- a. 360 degree shelter to two miles, shelter in downwind sectors to five miles,
- b. 360 degree evacuation to two miles, 360 shelter between two and three miles.
c. 360 degree evacuation to two miles, evacuation in at least three downwind sectors to five miles.
- d. 360 degree evacuation to three miles, evacuation in at least three downwind sectors between three and five miles.
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a SENIER REACTOR CPERATOR Peg 3 70 I QUESTION: 095 (1.00) I SELECT the event that is a safety Limit Violation,
- a. During a startup with two bypass valves full open, an APRM spike l causes a Div. 1 scram signal. The operator overreacts and inserts a full scram. All rods insert. Reactor pressure decreases 700- !
psig due to the subsequent cold water injection from feedwater and i the lack of dec:lc heat, j
- b. The reactor is at 55% power when a pressure regulator failure ,
causes the bypass valves to fully open. Reactor pressure : decreases to 700 psig and the MSIVs fail to close. Reactor power is 42% when the operator manually closes the MSIVs. The reactor autonatically scrams on the MSIV closure. Level is restored to normal with RCIC. ,
- c. Reactor power is 55% when both Recirculation Pumps trip. RPV !
Water level increases to +220 inches. The reactor fails to i automatically scram, but all rods insert when the operator manually scrams the reactor,
- d. Reactor power is 55% when the "A" RFPT trips. The reactor scrams on low level. HPCS and RCIC receive initiation signals but HPCS fails to start. RPV level decreases to +40 inches before RCIC is able to recover level. RPV pressure decreases to 700 psig with the subsequent injection.
i L l l l (********** END OF EXAMINATION **********) I l ;
SENIOR REACTOR CPERATOR P go 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 002 match with selected number in the blank a b c d , 003 a b c d 004 a b c d _ 005. a b c d. . 006 a b c d ' 007 a b c d 008 a b c d . 009 a b c d & P 010 a b c d 011 a b c d-012 a b c d __ 013 a b c d 014 a b c d ' 015 a b c d 016 a b c d . 017 a b c d 018 a b c d i 019' a- b c d , 020 a b c d l 021' a b c d ,
I L' SENIOR REACTOR CPERATOR. P093 2 [ \* ; it ANSWER S H E E~T o Multiple Choice (Circle or X your choice) j { If you change your answer, write your selection in the blank. { 022 a b c d 023 a b c d I
. 024 match with selected number in the blank a
b c t d
- 025 a b c d 026 c b c d 027 a. b c d 028 match with selected number in the blank a , ,
i b ' c d i ft9 a b c d ' 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
r .; l. (. SENIOR REACTOR CPERATOR.. Pcg3 3 t ANSWER SHEET i
; Multiple choice (Circle or X your choice)
If'you change your answer, write your selection in the. blank. 039 a b c d ! 040- a b c d ' 041 a b c d 042 a b c d l 043 a b c d I 044 a b c d ' 045 a b c d 046 a b c d j 047 a b c- d 048 a b c d 049 a b c d 050 a b c d t 051 a b c d 052 a b c d
-053 a b c d t
054- a b c d 055 a b c d 056' a b c d _ 057 a b c d 058 a b c d 059 a- b c d I
- 060- a b c .:
d 061 a- b c d ' 062- a b c d
- 063 a b- c d 4
SENIOR REACTOR OPERATOR - P;go 4 , ANSWER SHEET '! k Multiple Choice (Circle or X your choice) ) If you change your answer, write your selection in the blank. 064' a .b c d i 065 a b c d i 066 a b c d , 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 1 073 . a b c d ' 074 a b c d ~ 075 a b c d 076 a b c d 077 a b c d 078. a b c d 079 a b c d 080 a b c d 081 a b c d -
'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 087 a b c d 088 a b c d c
iSENIOR REACTOR OPERATOR p;g3 5-ANSWER- . SHEET l Multiple choice (circle or X your choice) j If you change your answer, write your selection in the blank. ( i 089 a b c d 090 a b c d i 091 a b c d 092 a b c d 1 093' a b c d ' 094 a b c d
- 095 a b c d P
1 (********** END OF EXAMINATION **********) i J > i
p_ i
;: SENIOR REACTOR OPERATOR t p;g3 71 t
' ANS**ER:' 001 -(1.00) i d.
i o
REFERENCE:
I PAP-205, " Operability of Plant Systems." [3.7/3.7) 294001K101 ..(KA's) ANSWER: 002 (2. 00) ' I
- a. 3 each response is 0.5 point
- b. 2.
- c. 4 ,
- d. 6
REFERENCE:
PAP-1701, " Miscellaneous _ Tagging." PAP-1401, " Safety Tagging." PAP-1402, " Control of Lifted Leads, Jumpers, Temporary Electrical , Devices,_and Mechanical Foreign Items."
< PAP-504, " Electrical operating Rules and Practices."
(3.9/4.5) 294001K102 ..(KA's) '
' ANSWER: 003 (1.00) b.
p @* L
r i .. 1 i SENIOR REACTOR OPERATOR Pcgi ?2 l
REFERENCE:
i PAP-1401, " Safety Tagging." [3.9/4.5) 294001K102 ..(KA's) , t ANSWER: 004 (1. 00) ; 9
- a. '
REFERENvE i PAP-118, "ALARA Program." PAP-514, " External Exposure Control."
'[3.3/3.8) +
294001K103 . . ( KA 's) ANSWER: 005 (1.00)- c. I
REFERENCE:
PAP-118, "ALARA Program." . PAP-514, " External Exposure Control." [3.3/3.8) 294001K103 ..(KA's) 1 ANSWER: 006 (1.00). ,
- d. '
k i
P0g3 73 SENIOR-REACTOR OPERATOR I. .
REFERENCE:
PAP-512, " Radiation Work Permits." (3.3/3.8) 294001K103 ..(KA's) ANSWER: 007 (1.00)
- j. b.
REFERENCE:
PAP-118, "ALARA Program." , (3.3/3.6) , 1 294001K104 ..(KA's)
' ANSWER: 008 (1.00)
- c. :
REFERENCE:
PAP-508, "PNPP Operating Rules and Practices." PAP-504, " Electrical Operating Ruleia and Practices."
.[3. 3/ 3 '. 6) i l
294001K107 ..(KA's) J 1' ? I . i l ANSWER - 009 (1.00) L C. l je t 3r
'SENI;R REACTOR CPERATOR P;g3 74 i
REFEPENCE: j PAP-516, " Confined Space Entry and Indusitial Hygiene Sampling."
-(3.2/3.4) ;
294001K114 ..(KA's) ; ANSWER: 010 (1.00) d.
REFERENCE:
a; PhP-1910, " Fire Protection Program." ' {3.5/3.8) 294001K116 ..(KA's) P ANSWER: 011 (1.00) c.
REFERENCE:
PAP-522, " Temporary Changes to Instructions." ,
'[2.7/3.7) ;
294001A103 ..(KA's) ANSWER: 012 (1.00) b. b c .- .
7-. p '. .
. SENIOR REACTOR-OPERATOR PC93 75
REFERENCE:
l t PAP-110, " Shift Staffing and overtime." (3.1/3.6) 294001A103 ..(KA's) i ANSWER: 013 (1.00)
- d. i i
REFERENCE:
SVI-C51-T0026, "APRM Plow Biased Power-Flow Verification" ; [3.1/3.6)- ;
~294001A108 ..(KA's) 5 ANSWER: 014 (1.00)
- b. I
REFERENCE:
Perry Technical Specifications sections 3.4.4 and 3.4.5. " PAP-1102, " Plant chemistry Control." [2.9/3.4) 294001A114 ..(KA's) LANSWER: 015 (1.00) . b. l l l
-SENIOR REACTOR OPERATOR Pag 3 76
REFERENCE:
EPI-A1, " Emergency Action Levels." EPI-A4, " Site Area Emergency." [2.9/4.7) 294001A116 ..(KA's) ANSWER: 016 (1.00) c.
REFERENCE:
EPI-A1, " Emergency Action Levels." [2.9/4.7) 294001A116 ..(KA's) ANSWER: 017 (1.00) a.
REFERENCE:
Perry. System Description Manual Chapter C22 OT-3036-C22-00, Obj. D [4.5/4.5)- 201001K205 ..(KA's)
' ANSWER: 018 (1.00) i a.
i i
i SENIOR REACTOR OPERATOR Pago 77 ;
REFERENCE:
Perry Technical Specifications 3/4.1.3.4. FTI-B02, " Control Rod Movements." . ONI-C11-2, " Uncoupled Control Rod." Perry System Description Manual Chapter C11 CRDM.- [
.[3.8/3.9) !
..(KA's) 201003K402 1
i ANSWER: 019 -(1.00) b. i
REFERENCE:
OT-3036-C11(RCIS)-01, Obj. G Perry System Description Manual, Chapter C11 , (3.5/3.5) 201005K403 ..(KA's) ANSWER: 020 (1.00)
- a. i
REFERENCE:
Perry Technical Specifications section 3.1.4.1. OT-3036-C11(RCIS)-01, Obj. J Perry System Description Manual, Chapter C11 Clinton LER 461-90-008 [3.2/3.2) , 201005K601 ..(KA's) b I
E' l
'SENIdR' REACTOR OPERATOR- Pega 78 ANSWER: 021 (1.00) b.
REFERENCE:
Perry System Description Manual Chapter B33 OT-3036-B33-01, Obj. G (3.5/3.6) 202001K505 . . (KA's) ANSWER:- 022 (1.00) c.
REFERENCE:
OT-3036-E12-01, Obj. G Perry System Description Manual Chapter E12 (3.3/3.4) 203000K402 . . (KA's) ANSWER: 023 (1.00) b.-
REFERENCE:
- OT-3036-E12-01, Obj. D ;
Perry System Description Manual Chapter E12. (3.3/3.5) 203000K604 . . (KA's) .
)
f SENI'R; REACTOR CPERATOR Pago 79 ANSWER: 024 (2.00)
.a. 3 Each response is 0.5 point.
b.' 4
- c. 2
- d. 4 i
REFERENCE:
OT-3036-G33/36-00, Objectives D, F. Facility examination bank question #3110. [3.6/3.6) 204000A303 ..(KA's) ANSWER: 025 (1.00) d.
REFERENCE:
l Perry System Description Manual Chapters E12, R23/24/25, B21 NS4 OT-3036-B21(NS4)-00, Obj. F, OT-3036-E12-01, obj. F (3.3/3.4)
.205000K601 ..(KA's) l -ANSWER: :026 (1.00) b.
1' 1 . i i
SENIOR REACTOR OPERATOR P093 80
REFERENCE:
OT-3036-E21-02 Obj. J OT-3036-B21C-00, Obj. E Perry System Description Manual, Chapter B21C ADS [3.7/3.7) 209001K105 ..(KA's) ANSWER: 027 (1. 00) d. REFERENCE PAP-606, " Condition Reports and Immediate Notifications." Perry Technical Specifications, Section 3.0.3, 3.5.1 [3.1/4.3) 209002G003 ..(KA's)_ ANSWER: 028 ( 2. 00)
- a. 3 Each response is 0.5 point.
- b. 1
- c. 2
- d. 4 REFERENCEt Perry System Description Manual Chapter C71 OT-3036-C71-02, Obj. J, K.
(3.4/4.3) 212000G006 ..(KA's)
r- > P g3 81 i SENIOR REACTOR OPERATOR . t l' ANSWER: 029 (1.00) , r
- c. ;
REFERENCE Perry System Description Manual Chapter C71. , OT-3036-C71-02, Obj. E. (3.2/3.4) o 212000K110 ..(KA's) i ANSWER: 030 (1.00) a.
REFERENCE:
'f OT-3036-C51(IRM)-01, Obj. E. I Perry System Description Manual Chapter C51 IRM ;
[3.3/3.3) i 215003A407 ..(KA's) ; ANSWER: 031 (1.00)
- d. i
REFERENCE:
'OT-3036-C51(IRM)-01, Obj. E. OT-3036-C51(SRM)-01, Obj. F. -
Facility examination bank question #3106 _(3.3/3.3) 215004K303 ..(KA's) f
I ' 'J P0g] 82 I SENI R REACTOR OPERATOR' l ANSWER: 032 (1.00) c.
REFERENCE:
OT-3036-C51(PRM)-03, Obj. K L Perry System Description Manual Chapters B33 and C51 PRM. u ,[3.6/3.6) 215005K109 ..(KA's) [ ANSWER: 033 (1.00) c.
.REFERENFE:
OT-3036-E51-00, Obj. D. Perry System Description Manual Chapter E51. (4.0/4.3) 216000K303 ..(KA's) ANSWER: 034 (1.00) d.
REFERENCE:
OT-3036-E51-00, Obj. F
< Perry System Description Manual Chapter Ef1
[3.9/3.9) l 217000A402 ..(KA's)
7 . I p 8ENIOR REACTOR OPERATOR' pag 3 83 ' I & ANSWER: 035 (1.00) >
- a. ,
REFERENCE:
OT-3036-B21C-00, Obj. E Perry System Description Manual Chapter B21C ADS [3.1/3.3) i t 218000K201 ..(KA's) ANSWER: 036 (1.00) ' i d.
REFERENCE:
ONI-G41, " Loss of Reactor Cavity / Fuel Storage Pools Level." f
.[3.1/3.1) 233000G014 ..(KA's)
ANSWER: 037 i (1.00) ;
- c. '
REFERENCE:
-Perry System Description Manual Chapter F15.
SYS-5014-F15-00', Obj. E SOI-F11/F15, " Fuel Handling, Refueling, and' Auxiliary Platforms." [3.1/3.7) > 234000K502_ ..(KA's) 1
?
Pago 84 \'[SENIORREACTORE}PERATOR { J ANSWER: 038 (1.00) f I b.
REFERENCE:
i Perry System Descrip9. ion Manual Chapter D17A PRM OT-3036-D17A-00, Obj. D , OT-3036-B21(NS4)-00, Cbj. G ; (3'6/3.6) 239001A105 ..(KA's) i ANSWER: 039 (1.00) d.
REFERENCE:
i f Perry System Description Mar;ual Chapter B21/N11 OT-3036-B21/N11-01, Obj. C ( (3.9/3.9) ;
'239002A309 ..(KA's)
ANSWER: 040 (1.00) :
- c. ,
REFERENCE:
1
'OT-3036-N32/C85-00, Objective Perry System Description Manual Chapter N32/C85 (3.3/3.3)
;241000A116 ..(KA's) i
Q* SENIOR' REACTOR OPERATOR Pago 85
,.t ANSWER: 041 (1.00) i
- b. s
REFERENCE:
OT-3036-N32/085-00, Obj. K ONI-C85-2, " Pressure Regulator Failure Open" 6 (3.5/3.7) ! 241000A201 ..(KA's)
. ANSWER: 042 (1.00) ,
a.
REFERENCE:
Perry System Description Manual Chapter N21, Chapter C11 CRDH OT-3036-C11(CRDH)-02, Obj. B. (3.1/3.1) 256000K105 ..(KA's) ANSWER: 043 (1.00)
.d.
REFERENCE:
Perry System Description Manual Chapter N21/N61. OT-3036-N21/N61-01, Obj. C, E. , (3.1/2.9) 256000G010 ..(KA's) , L l8
! SENIOR' REACTOR OPERATOR POg3 86 L ,
ANSWER: 044 (1.00) d .-
REFERENCE:
Perry System Description Manual, Chapter N27/N27A. OT-3036-N27-01, Obj. E (3.3/3.5) 259001A301 ..(KA's) ANSWER: 045 (1.00) a.
REFERENCE:
ED, HAVE FACILITY CHECK FOR TECHNICAL ACCURACY ! OT-3036-N21-01, Obj. F. Perry System Description Manual Chapter N27. SOI-N27, "Feedwater System." (3.2/3.3) 259001G010 ..(KA's) I LANSWER: 046 (1. 00) c. i I ( .
r- - O SENIOR-REACTOR OPERATOR pega.87 i l
REFERENCE:
l 1 Perry System Description Manual Chapter B33
'OT-3036-B33-01, Obj. F, G (3.0/3.1) l l
259002K401 ..(KA's) !
. -i ANSWER: 047 (1.00) d.
REFERENCE:
t Perry System Description Manual Chapter C34 ' OT-3036-C34-02, Obj. C O!!I-C34, "Feedwater Flow Control Malfunction" , (3.1/3.1) l 259002K603 ..(KA's) i ANSWER: 048 (1.00) 1
- b. ;
REFERENCE:
OT-3036-R10-00, Obj. C, E Perry. System Description Manual Chapter.Rio (3.1/3.2] , 262001A201 ..(KA's) ANSWER: 049 (1.00) l b..
, - ~ ~ - a
F , POg3 88 L' SENIOR REACTOR OPERATOR
REFERENCE:
e Perry Technical Specifications section 3.8.2.1 '
-0T-3036-R42-03, Obj. G (3.1/3.8). i 263000G005 ..(KA's) i f ' i ANSWER: -050 (1.00)
- b. .
t
REFERENCE:
OT-3036-R43/48-01, Objectives C, H. Facility examination bank question #3092 Perry System Description Manual, Chap. R43 : (3.7/3.5) 264000X402 ..(KA's). Ai!SWER: 051 (1.00) , a. l
REFERENCE:
1 OT-3036-R43/48-01, Obj. F Perry' System Description Manual Chapter R43 > 1 (3.4/3.,4)
.264000K505 ..(KA's) .
ANSWER: 053 (1.00) b. I 1 l' L l h
P0g3 89 SENIOR REACTOR OPERATOR i
. 1
REFERENCE:
~OT-3036-N64-01, obj. E, K, OT-3036-D17A-01, Obj. D l Perry System Description Manual Chapter N64 (3.0/3.2) 271000K602 ..(KA's) i 4
ANSWER: 053 (1.00) -
- b. ;
REFERENCE:
Perry System Description Manual Chapter D17A i OT-3036-D17A-00, Obj. D (3.4/3.6) 272000K306' ..(KA's) P v ANSWER: 054 (1.00)
- b. -
i
REFERENCE:
i 4 Perry System Description Manual, Chapter M25/M26 OT-3036-M25/M26-00, Obj. B # (3.3/3.5) 290003A301 ..(KA's) f 4 ANSWER: 055 (1.00) d.
^
SENI;R REACTOR OPERATOR Pag 3 90 :
REFERENCE:
ONI-B33-2, Loss of one or Both Recirculation Pumps , Facility examination bank question #3072 (3.5/3.8) , 295001A201 ..(KA's) P ANSWER: 056 (1.00) l 1 b.
REFERENCE:
i ONI-B33-2, " Loss of One or Both Recirculation Pumps." , Perry Technical Specifications, section 3.4.1.1 i (3.2/4.1) 295001G003 ..(KA's) LANSWER: 057 (1.00) r
- d. .
REFERENCE:
ONI-N52, " Loss of Main Condenser Vacuum." [3.2/4.1)
.295002K201 ..(KA's)- .
ANSWER: '058 (1.00)
- b. ,
6 1
3 ,, , -7 -, .4
'! SENIOR REACTOR. OPERATOR Page.91 c
REFERENCE:
f 1-ONI-RIO, " Station Blackout."'
.(4.2/4.3) +
295003A102- .!(KA's)
. AlPFER:- 059: (1.00) a.
REFERENCE:
'ONI-R4'2-4,." Loss of DC Bus D-1-A."
n'- , [3.2/3.4)
- 295C04C010- ..(KA's) i lANSWER:- '060 ( 1". 0 0 )
;c.
REFERE.::CE: fi ,
' Perry System Description Manual Chapter;N31/11A/39
-(3'.4/3.5)-
~295005A207 ..(KA's) ,
)
' ANSWER: -061- (1.00)
',$l ' h g
(( jI' a y, I . p
+ i5
'i
- YSEN OR REACTOR OPERATOR' -P293-92 r
~!
~ REFERENCES-PEI-B13, " Reactor Pressure Vessel Control."
SOI-C11(CRDH), " Control, Rod = Drive Hydraulic System." 3 (3.7/3.8]
-295006K003 ..(KA#s)
,-ANSWER: 062- (1.00) ,
i (
- .b. .
REFERENCE:
Perry system Description Manual Chapter-E22A,.B21 NBPI.
~(3.4/3.5]
295008A106 -..(KA's)
' ANSWER: -063 (1.00) 7 a.
i
REFERENCE:
PEI-D23-3, "DrywelliTemperature Control." PEI-D23-1, " Containment Temperature control." ll3.4/3.5) 1 295010A101- ..(KA's) t I~ i
' ANSWER: 064 (1.00) d.
l 1 l
M . [ J- SEhlbRREACTOROPERATOR pag 3 93-2 . e
, "L.;
REFERENCE:
PEI-B13,=" Reactor Pressure, Vessel Control."
- 1 PEI-D17,." Radioactivity Release Control."
s PEI-G42,L" Suppression; Pool Level Control."~ ( 4 '. 2/4 . 5 ) .. 1 "^ '
~
.295010G011- :..(KA's) s ,
t ANSWER:- 065 (1.00) l e ,
?
b.
REFERENCE:
,lPEI-D23-1, " Containment Temperature Control."
~ [ 3 . 6/ 3 '. 9 )
3. J295011A201 ..(KA's). t ANSWER:i . 066L (1.00)- ' l y a f 1C.
- d. . 7
REFERENCE:
e
-ONI-C85-1, " Pressure Regulator Failure-Closed."
.m :-[ 3.~3/ 3. 4 )
,g 295014A106 ..(KA's) m .
J, i; EANSWER:L , . 067, . (1.~ 0 0) 1
- b. .:
g'j 1 g i t- ' , , ' .
-. q .
l
SENIOR: REACTOR; OPERATOR.
'i g, pag 3 94 m4
; REFERENCE ;
e i ONI-B13,1 " Reactor-Pressure Vessel Control." [4'.1/4.3) 295015A201 ..(KA's)'
~
~
-e i 4
ANSWER:' 068 ( l'. 0 0 ) t d. i
REFERENCE:
s PEI-B13, " Reactor Pressure Vessel Contr'1", step.3.1.6. ( [4.0/4.1) 295015K204' i
..(KA's) '
i
" ANSWER: - 069 (1.00) i r
a.- ; i
REFERENCE:
ol ONI-C61:
.[3.8/3.8)
- 295016G010 ..(KA's)
~
, -i
-ANSWER: 070 (1.00) i 8..
4 i
, r
?
4 .__ ___
SENIOR REACTOR ~ OPERATOR Pcga 95-
REFERENCE:
i EPI-Al', " Emergency Action' Levels."
"J J PAP-606,-" Condition Reports and Immediate Notifications."-
LEPI-A4, " Site-Area: Emergency." . (3.1/4.7]- [' 295017G002 . . ( K?. 's )
' t!
[ ANSWER: 071 (1. 00) i a.
?
REFERENCE:
ONI-P43, " Loss of Huclear Closed-Cooling Water."
.[3.4/3.3) '
')'
4 29S01BG010 ..(KA's) - c
' ANSWER. 072 (1. 00) ,
b.
REFERENCE:
u_
.ONI-P41,I" Loss of Service Water."
ONI-P43, " Loss of. Nuclear closed Cooling 1 Water." ' ONI-P44,'" Loss of Turbine Building. Closed Cooling Water."
.ONI-P52, " Loss of Service and/or Instrument. Air." .
.[3.8/4.1) -
295018G011 ..(KA's) ' i '; f(ANSWER:. 073, (1.00)
+
d. I 1
g -u 1 3. F , .
- g. - ;;; SENIOR! REACTOR OPERATOR L pgg3 96'
REFERENCE:
~
ONI-PS2, " Loss of L Service and/or Instrument Air. "' [3.9/4.1) 295019.G010 ' ..(KA's) ANSWER:' 074 (S. 00)
.b.
L
REFERENCE:
- ONI-E12, " Loss:of Shutdown Cooling."
, (3.7/3.7) I b '95021A104 ..(KA's)-
.g ANSWER:, 075 (1.00) o ;b. , ,
. -J l
REFERENCE:
I p 4 o
>i.
ONI-E12-2,." Loss of Shutdown' Cooling." * , Ferry System Description Manual Chapter-E12. (.. . [ 3. 9/3 . 9 ] . u j o 295021K102 ..(KA's) -l sl o NSWER:!- 0761 (1.00) ' ]
, :a.' '
.j
!! ! [
~
h
~!
c_ V: k A
,} ,,
, 1 F
q SENIOR REACTOR OPERATOR pago:97 [RbFERENCE ONI-C11-1,:" Inability tofMove Cchtrol~ Rods." Perry Technical Specifications'section 3.1.3.3. !
-[3.7/3.6)'
295022G010. ..(KA's) ! i
. ANSWER: 077 (1.00)
[ b; '
REFERENCE:
=1 OT-3036-M14-00,oObj. D
-Perry System Description Manual Chapter.D17 PRM, M14 j
ONI-J11-2, " Fuel: Bundle Rupture During Fuel Handling." t (3.3/3.4] ' 295023A108 . .~ ( KA ' s ) s l ANSWER: 078 (1.00)- ,
- a. ,
REFERENCE:
, OT-3034-02-D23-2-00, Obj.-C.
PEI-D23-2, "Drywell'and. Containment Pressure Control."
, [3.6/3.6) .
295024A118 ..(KA's) . 1 ANSWER: . "'9' (1.00) m : b. 4.
.f I r \
# I t
l J i
r;c m 'i i
. n R [ SENIOR. REACTOR OPERATOR; k
Pag 3'98 t
REFERENCE:
Perry: Technical'- Specifications. section 6.7.1, 2.1.3.
- (4.4/4.7): .
.' 5-295025X105 ..(KA's) !
t I g ANSWER:. 080 (1~. 0 0) - ,.I 1
- c. .
o I REFERENCEi l
' e larry System Description Manual Chapter C22.
[3.'3/3.7)
- t 295025A108 ..(KA's)
, .i
-A SWER:- 0810 ' (1.00) '
i D.. t
REFERENCE:
\0NI-B21-1, "SRV Inadvertent Opening / Stuck Open."
Perry Technical Specifications section.3.6.3'.1. .
. i, [4.1/4.2]- ,
$ '295026A201 -..(KA's) o l r (ANSWER:; -082 (1.00)
. m. . .;
Y['
! d. . .
.p !
1, !
'l . t i s > ,
l ' )4 l g l'h-(l ;' xi' , , s i
.[ j,^ : 5-L T .'iv ,1.!..
l ~:. J ']i l , , , . . . . , ,
t y> , b I ' SENIOR'. REACTOR 10PERATOR . Pago 99
.i
.}
REFERENCE:
=PEI-D23-2,;"Drywell Temperature Control."' .
. [ 3. 0/3. 2 ) . 7
..(KA's).
a 295027K102--
- LANSWER:'
083 (1.00) t a.-
REFERENCE:
OT-3034-02-D23-1, Obj. B. (3.7/3.8) 295027K301 ..(KA's) 1 ANSWER: 084' (1.00)
.u c.
REFERENCE:
h,',o ~ OT-3034-02-D23-3-00,,Obj. C.
. [3.5/3.7)-
M 295028K101 ...(KA's) 10 i ANSWER: '085 :(1.00) - \ , I y a .- 4
- i. k p.
L . i- . _ - . _ , .. _
i E
}
SENIOR.. REACTOR OPERATOR-
, , _. Paggloo; t
r
REFERENCE:
, t
') ' ' s PEI-G42, " Suppression Pool' Level Control."' -i -
' [ 3. 5/3. 9 ) .
295029K301 ..(KA's)'
; ~
. ;r
' ANSWER: 086 (1.00) '
a . -- . .. r o > '
REFERENCE:
PEI-B13, " Reactor Pressure Vessel-Control." 1 [3.6/349) -
- 295030G007 ..(KA's)
. ANSWER:
087, (1.00) Com E
REFERENCE:
lPerryJTechnical Specifications section 3.6.3.1. [3.4/4.4):. 295030G008 ..(KA's)1 + _"k IANSWER: . 088 (1.00)
' e 4, . b..
p b t 4 1 1 ' u , 't.- J
- l8
.i l
l
>>" ..a ,
J
n
?v
' - aySENIOR(REACTOR OPERATOR 7 i PEgo10h ,. it ' U REFERENCE': ' PEI-B13, " Reactor Pressure Vessel Control." [3.8/3.9): '
'295031A108- ..(KA's)-
'089-ANSWER: (1.00)
- i
- d. ,
/
REFERENCE:
-[4.6/4.8)
;7; 295031A204 ..(KA's)'
. ANSWER:. 090 - (1. 0 0 ) ,
- b . :-
1
REFERENCE:
PEI-B13,," Reactor Pressure' Vessel Control." OT-3034-02-ATT5-00, Obj. C. 'l
'[3.7/4.1) 1 295031K103 ..(KA's) 1
' q ANSWER: ' 091 = (1. 00) cf .
. c.
i!! [1 . i l g. lf . m<- .
U f SENIOR REACTOR: OPERATOR > Pago102
REFERENCE:
r Perry Technical ~ Specifications section, 3/4.6'.6.1, B3/4.6.6. !
'(3'.7/4.2]; - -
-- t 295035K102'- ..(KA's)' .
i ANSWER:- 092 (1. 0 0) ~
. l
- d.
r
REFERENCE:
i PEI-B13, " Reactor Pressure Vessel Control." .
^
[3.9/4.6) 295037G012 ..(KA's) .o.
; ANSWER:' 093 t1.00). 'I
'b. ..
y
REFERENCE:
;- t -
PEI-B13, " Reactor Pressure Vessel' Control." M .[.4'.'2/4.3)- 295037K307 ...(KA's) 4 M i ANSWER: 09.4 (1.00) i m . 1 I -b '
!h,.
i t- 1 L L e
+ , + ,- w -
, , - - - . , a +
LSENIOR-REACTOR ~ OPERATOR peg 31031
REFERENCE:
EPI-B8,, Protective ~ActionsLand Guides."
-[ 3. 3/4. 3 ) .
295038A203- ..(KA's)
. s ,
e ANSWER: 095 (1.00) , f b.
?
REFERENCE:
i Perry Technical' Specifications,1section 2.1.1. [3.8/4.4) j 295006G003 ..(KA's)
~
h
~f i
m; d i k - f i i s l (********** END OF EXAMINATION **********) a
-}}